DE102017111935A1 - Method and device for multi-element analysis based on neutron activation and use - Google Patents

Abstract

The invention relates to a method for multi-element analysis based on neutron activation, comprising the steps of: generating fast neutrons with energy in the range of 10 keV to 20 MeV; Irradiating a sample (1) with the neutrons; Measuring the gamma radiation emitted from the irradiated sample to determine at least one element of the sample; wherein according to the invention, the irradiation of the sample is carried out unpulsed in a continuous manner, wherein the measuring takes place during the irradiation. This expands the analysis and provides a flexible process. Furthermore, the invention relates to a corresponding device and the use of a detector unit for the multi-element analysis.

Description

TECHNICAL AREA

The present invention relates to a method for multi-element analysis based on neutron activation by irradiating a sample with neutrons. Furthermore, the present invention relates to a corresponding device, in particular with at least one collimatable detector. Last but not least, the present invention also relates to the use of a controller or a computer program product therefor. In particular, the invention relates to a method and an apparatus according to the preamble of the respective independent claim.

BACKGROUND OF THE INVENTION

In many areas of industry, the analysis of substances or materials, particularly with regard to their elemental composition, is of great importance, in particular with regard to dangerous goods or wastes or recycling materials or raw materials or the quality control of semi-finished products or industrial products. One of the analysis methods carried out so far is the so-called multi-element analysis, by means of which individual elements of a sample are determined, without first having to know the exact composition of the sample.

The multi-element analysis can be carried out by neutron activation or, for example, by means of X-ray fluorescence analysis or mass spectrometry. So far, multi-element analysis by means of neutron activation has been carried out by irradiation according to specific time specifications. It has been shown that a good informative value of the measurement results can be ensured if, in the case of pulsed neutron irradiation, prompt gamma radiation is evaluated after a certain time window as a function of the manner of the pulsed irradiation. The time window for the detection of the gamma radiation is started after a waiting time, which follows at the end of each neutron pulse, and ends before the next neutron pulse is emitted.

The WO 2012/010162 A1 and the DE 10 2010 031 844 A1 describe a method for the non-destructive elemental analysis of large-volume samples with neutron radiation and an apparatus for carrying out the method. In the method, the sample is pulsed with fast neutrons, wherein the gamma radiation emitted by the sample is measured after a certain time window after a neutron pulse before a new neutron pulse is emitted to the sample. The gamma radiation is measured between the individual neutron pulses. The amount of an element contained in the sample is evaluated after subtracting the background signal generated by the sample and / or by the measuring system from an area content analysis of a photopeak. In this case, the determination of the neutron flux takes place in connection with a metallic coating of the sample (neutron flux monitor). The measuring method is also based on the knowledge that the measurement can be made possible by a moderation process and by observing a time window after a respective neutron pulse. The detection in the detector of induced gamma radiation, resulting from inelastic interactions, can be filtered out due to the time window after a respective neutron pulse and thus be masked out during the measurement. As gamma radiation, prompt gamma radiation is evaluated.

Also EP 1 882 929 B1 and WO 01/07888 A2 describe methods in which neutrons are pulsed on the sample and a time window is maintained after each pulse until the prompt gamma radiation emitted by the sample is measured. Comparable methods are eg also in EP 0 493 545 B1 and in DE 10 2007 029 778 B4 described.

The object is to provide a method and a device with which the multi-element analysis of samples can be facilitated by means of neutron activation. It is also an object to design a method and a device for multi-element analysis by means of neutron activation in such a way that results in a broad field of application. Last but not least, it is an object to provide a flexible as possible non-destructive method and a corresponding device for multi-element analysis based on neutron activation with a high quality or with a very reliable, secure way of measuring and evaluating emitted gamma radiation, even in the event that to be examined sample with respect to the element composition and the sample geometry should be difficult to analyze and / or a destructive or influencing partial sampling of the sample to be examined is not desired.

SUMMARY OF THE INVENTION

At least one of these objects is achieved by a method according to claim 1 and by a device according to the independent device claim. Advantageous developments of the invention are explained in the respective subclaims. The features of the exemplary embodiments described below can be combined with one another, unless this is explicitly denied.

A method for multi-element analysis based on neutron activation is performed with the following steps: generating fast neutrons with energy in the range of 10 keV to 20 MeV; Irradiating a sample with the neutrons; Measuring the gamma radiation emitted from the irradiated sample to determine at least one element of the sample. According to the invention, it is proposed that the irradiation of the sample takes place in an unpulsed, continuous manner, the measurement taking place irrespective of the irradiation time (irrespective of the time course of the irradiation), in particular without time intervals given by neutron pulses, during the irradiation, in particular simultaneously for the irradiation over the same period of time like the irradiation, in particular continuously during the irradiation.

As a result, a non-destructive multi-element analysis based on neutron activation can be provided for diverse types of samples and with high metrological flexibility and robust, reproducible, reliable results. The sample is continuously irradiated with neutrons without individual pulses, for example continuously over a period of several seconds or minutes or hours, whereby the gamma radiation emitted / emitted by the sample can be measured simultaneously with the irradiation. It has been shown that the neutrons can be generated in particular by means of a generator for the fusion of deuterons (deuterium nuclei), in particular with deuterium gas as a gaseous target or as a fuel. The present invention enables measurement and evaluation based on comparatively low-energy irradiation over a long period of time, whereby the analysis can be carried out very accurately and reproducibly.

According to the state of the art, it has hitherto been necessary to irradiate in a pulsed manner, so far a waiting time after a respective neutron pulse has been required. The pulse length used to be usually in the range of ten to several hundred microseconds (μsec.). In contrast to pulsed irradiation, both the prompt and the emitted gamma radiation emitted by the sample are measured simultaneously with the irradiation, the energy resolution of the detector allowing the division into prompt and delayed gamma radiation. In contrast to pulsed irradiation, no time window (previously usually at least 5 μsec.) Must be observed until the measurement / evaluation of emitted radiation can take place. There is no waiting time for the start of the detection of gamma radiation necessary. A temporal coordination with the end of a respective neutron pulse is no longer required, but it can be continuously irradiated and measured. This also makes it possible to reduce the measuring time for the analysis of the samples.

The measurement can be done completely without time window, or optionally partially with time window. At least part of the emitted gamma radiation is preferably at least time-independent measured without time window. The method may include moderating the fast neutrons, at least temporarily.

The simultaneous measurement of the gamma radiation allows a high performance of the measuring system. In a standard mode of operation, both prompt and delayed gamma radiation can be measured, with a focus on prompt gamma radiation. The measurement simultaneously with the irradiation can take place in a continuous manner, in particular with the same timing as the irradiation, or in individual time windows irrespective of time specifications for the irradiation. For example, a continuous irradiation, but optionally a measurement takes place only at small intervals. Measuring at the same time as continuous irradiation means that no time windows have to be taken into account, but that the measurement and evaluation can be carried out in a very flexible manner and both types of radiation, ie prompt and delayed gamma radiation, can be evaluated. As a special feature of the present invention, the measurement / detection of the gamma radiation can be emphasized independently of temporal relationships in the neutron irradiation.

According to the prior art was pulsed irradiated, which previously required a waiting time after each neutron pulse. The pulse length used to be usually in the range of ten to several hundred microseconds (μsec.). Immediately after a respective neutron pulse, the background signal of the previously used measuring systems was too high, so that the signal to noise ratio (SNR) directly after the neutron pulse was too poor to be able to evaluate the measurement. Therefore, no meaningful spectrum of gamma radiation could be detected. In previous ones The procedure had to allow the neutrons a certain time window to make the measurement, especially after the neutrons were emitted. Usually, this time window is at least 5μsec. In this time window, the probability of interactions in the sample increases, so that after a certain waiting time (or moderation time) after a respective neutron pulse with a sufficiently good signal to background ratio SNR could be measured. The data acquisition took place in a time-shifted manner as a function of the neutron pulses.

On the one hand, the gamma radiation measured and evaluated according to the present method is, on the one hand, prompt gamma radiation which is emitted immediately after an interaction of the neutrons with the atomic nuclei of the sample. The time to emission is about 10exp-16 to 10exp-12 seconds in the case of prompt gamma radiation, which period is so short that immediate / immediate emission can be used. In the case of prompt gamma radiation, there is no metrologically relevant time delay between neutron capture and emission of the gamma radiation. On the other hand, delayed gamma radiation is also affected, which is emitted at the disintegration of the activated atomic nuclei with a time delay according to the characteristic half-life. Delayed gamma radiation is emitted out of the atomic nucleus after a neutron capture according to the characteristic half-life of the formed radionuclide. In classical neutron activation analysis (NAA), the neutron capture cross section and the half-life of the activated radionuclide have an influence on the emitted radiation. According to the invention, two measurement concepts can be interconnected: classical neutron activation analysis (NAA) on the one hand, and a prompt gamma NAA on the other hand (PGNAA). In this case, a distinction can be made between prompt and delayed gamma radiation based on the energy of the gamma radiation (in particular the position of the maximum of a peak) and the energy resolution of the detector.

As expected, the greatest gain in knowledge can be achieved by evaluating prompt gamma radiation. However, there are a whole series of elements, such as Lead, which does not give a good prompt signal. Therefore, in many applications or numerous types of material samples, it is useful to evaluate both prompt and delayed gamma radiation. Optionally, the irradiation may optionally also be in a pulsed manner, at least temporarily, to measure only delayed gamma radiation. Optionally, regardless of the manner of irradiation, only delayed gamma radiation may be measured, especially in an analysis for lead.

The gamma radiation emitted from the sample is measured in an energy-resolved manner in one or more detectors. This results in a measured gamma spectrum, in particular according to a record of the number of events detected in a gamma ray detector as a function of energy. The energy of the gamma radiation is used to identify the elements of the sample. By means of the measured energy-dependent radiation intensity, the quantification of an element mass takes place. The mass fraction of an element contained in the sample, after subtracting the background signal, is evaluated from the area of the photopeaks which the element produces in the gamma spectrum. Since the irradiated elements in the sample usually emit gamma radiation at different energies, all evaluable gamma energies of an element are taken into account in the analytical evaluation for the mass determination and in the uncertainty analysis for the mass determination. It has been shown that the analysis based on all evaluable gamma energies of an element has the advantage that a broad data basis can be used and that plausibility checks can be carried out. Furthermore, a remaining measurement uncertainty can be reduced, and the accuracy of the measurement method can be increased.

The analytical evaluation for mass determination in the multi-element analysis is based in particular on the calculation of energy-dependent photopeak efficiencies of gamma emissions from the sample and from individual partitions of the sample and the calculation of the neutron spectrum and the neutron flux within the sample and within partitions of the sample. For these calculations, initial assumptions about the elemental composition can be made, which are derived from the evaluation of the gamma spectrum. It has been found that the results of the multi-element analysis can clarify the a priori made initial assumptions for the calculation of photopeak efficiencies, neutron spectrum and neutron flux and also increase the accuracy of the measurement method, so that the method preferably carried out iteratively with respect to the composition of the sample so long until the calculated composition of the sample stabilizes. The non-destructive method for multi-element analysis based on neutron activation is possible by this type of analytical procedure in an automatic manner, in particular iteratively, with as input parameters in particular only the shape of the Specimen and the mass and the neutron source strength are required. The neutron source intensity can be obtained from the measuring system or device as a controlled variable.

By means of this method for multi-element analysis based on neutron activation and a device for carrying out the method, it is possible to examine various samples in a non-destructive manner in a simple way with respect to the material composition. Examples of analyzable samples include soil samples, ashes, water samples, sewage sludge, electronic scrap, chemotoxic or radioactive waste. The analysis of samples can be done in batch mode or online at a mass flow. The analysis of samples may be performed for purposes such as quality assurance, targeted sorting, process control and / or verification.

In comparison to previous methods, the method according to the invention is characterized in particular by the following properties: continuous emission of fast neutrons; Continuous measurement of gamma spectra; Collimated measurement of the entire sample or of individual partial volumes of the sample (partitions); in particular, the sample is irradiated with 2.45 MeV neutrons (<10 MeV, comparatively small starting energy); Gamma radiation is evaluated by evaluating the signal of each partition; the analytical evaluation for the determination of the element masses takes place in particular based on the simplified assumption that the element mass is homogeneously distributed in a partition; and / or there is a rotation and axial displacement of the specimen to the detector. After or during the emission of fast neutrons, the fast neutrons can be moderated in a moderation chamber, in the sample chamber and / or in the sample itself until they are sufficiently thermalized.

In contrast, the analysis has hitherto usually been carried out by a method having the following properties: pulsed irradiation with fast neutrons; Measurement of the gamma spectra in predefined time intervals or time windows after a respective neutron pulse or between the individual neutron pulses; the sample is measured integrally without defining a collimator; the sample is in particular irradiated with 14.1 MeV neutrons (> 10 MeV); there is a rotation of the specimen in front of the detector, and the gamma radiation is measured as a function of a rotation angle of the irradiated sample; the analytical evaluation for the determination of inhomogeneously distributed element masses is based on the simplified assumption that the element mass is punctiform. The integral neutron flux in the sample can be determined by means of a metallic coating of the sample.

The following will briefly discuss the terms used in connection with the present invention.

As a shielding is preferably a material or a unit to understand what / which encloses the device or measuring system and reduces the gamma and Neutronenortsdosisleistung outside the measuring system. The irradiation or irradiation is preferably to be understood as operation of a neutron generator and generation and emission of neutrons onto at least one sample in order to bring about the emission of characteristic gamma radiation from the sample for the elemental composition.

The detector unit is preferably a unit or module of the measuring system, comprising one or more detectors, with which detector unit the gamma radiation emitted from the sample or from individual partitions of the sample is measured in high-resolution. A respective detector may have an extension of e.g. 5 to 10cm in one spatial direction.

A collimator is preferably a unit or module of the measuring system which limits the field of view of a detector to a spatial area with a higher detection probability for gamma radiation. Collimation may also be specific to individual segments / partitions of the sample. As a measuring system is preferably a metrological system for generating ionizing radiation to understand the purpose of the multi-element analysis of samples. The device described here can be referred to in one embodiment as a measuring system.

As moderation chamber is preferably an assembly of the measuring system for the moderation of neutrons to understand, in particular by means of graphite or at least partially made of graphite. Depending on the desired manner of measurement / evaluation, the moderation can optionally be provided in the sample chamber, and / or in a separate moderation chamber.

A neutron generator is preferably an assembly of the measuring system which emits fast neutrons (in particular 2.45MeV neutrons or furthermore also generally neutrons with <10 MeV) and disposed within the shield. The neutron generator may optionally be surrounded by a moderation chamber provided separately from the sample chamber.

The neutron flux is preferably a product of the neutron density (free neutrons per cm 3) and the mean amount of the speed of the neutrons (cm / sec).

The neutron spectrum is preferably to be understood as the relative distribution of the neutron energy over the entire energy range of the neutrons.

A partition is preferably to be understood as a predefinable / predefined spatial area within the sample, the sum of all partitions resulting in the entire sample or defining the entire sample body.

A preferred number of partitions may be chosen depending on the size of the sample and the metrological task, for example between 1 and 60 partitions. The volume of a partition can range from a few cubic centimeters to liters. For particularly small samples, e.g. a few cubic centimeters, it may be advantageous to define only a single partition.

The photopeak efficiency is preferably a detection probability for the complete energy deposition of a gamma emission in the detector.

As gamma emission gamma radiation can be understood regardless of their energy level. Specific gamma radiation has a specific energy. Gamma emission as such is the reaction after irradiation with neutrons. The analysis is therefore carried out specifically with respect to individual types of gamma radiation from the spectrum of gamma emission. The gamma emission spectrum is used to detect signals of prompt and delayed gamma radiation.

As the sample, it is preferable to understand a solid or liquid amount of material which is selected for analysis and is the subject of the investigation, e.g. including soil samples, ashes, water samples, sewage sludge, chemotoxic or radioactive waste.

The sample chamber is preferably an assembly of the measuring system in which the sample is placed during the irradiation, and in which the sample can optionally also be displaced, in particular during the irradiation.

As a sample carrier is preferably an assembly of the measuring system to understand, which receives the sample and which is arranged in the sample chamber. By means of the sample carrier, a local displacement of the sample can take place.

In the following, the method according to the invention will first be described in general terms, with details of the method being discussed in detail below.

Through the operation of one or more neutron generators, a sample within a measuring system is continuously irradiated with neutrons and measured simultaneously to the irradiation induced by the neutron interactions / emitted gamma radiation.

As already mentioned, the gamma radiation is, on the one hand, prompt gamma radiation emitted immediately after interaction of the neutrons with the atomic nuclei of the sample and, on the other hand, delayed gamma radiation emitted upon decay of the activated atomic nuclei according to the characteristic half-life. The gamma radiation emitted from the sample can be measured in an energy-resolved manner in one or more detectors. This results in a measured gamma spectrum, per detector. The gamma spectrum is the recording of the number of events detected in a gamma detector as a function of energy. The energy of the gamma radiation is used to identify the elements of the sample. By means of the measured energy-dependent radiation intensity, the quantification of an element mass can take place.

Calculation of element masses for partitioned and non-partitioned measurements

wobei

• - (P_R)_{E γ} die Netto-Zählrate des Photopeaks des Elementes für die Gamma-Energie E_γ ist,
• - M die molare Masse des entsprechenden Elementes ist,
• - N_A = 6.0221408 ▪ 10²³ die Avogadro Konstante ist,
• - x die Position in dem Probevolumen V ist,
• - E_n die Neutronenenergie ist,
• - m(x) die Verteilungsfunktion für die Masse des entsprechenden Elementes im Probevolumen ist,
• - ε_{E γ} (x) die Photopeak-Effizienz für Gammastrahlen des Elementes an der entsprechenden Energie E_γ, die an der Position x emittiert worden ist,
• - σ_γ(E_n) der partielle Gamma-Produktions-Wirkungsquerschnitt ist, und
• - Φ(x, E_n) der Neutronenfluss als Funktion der Position und Energie in der Probe ist.

The mass fraction of an element contained in the sample, after subtracting the background signal, is calculated from the area of the photopeaks produced by the element in the gamma spectrum. The at The net photopeak count registered by the multi-element analysis is dependent on the following parameters of influence, which relationship was particularly considered in the following publication: GL Molnar (Ed.), Handbook of Prompt Gamma Activation Analysis with Neutron Beams, Kluwer Academic Publishers, ISBN 1-4020-1304- 3 (2004). $${\left(P_R\right)}_{e_{\mathrm{\gamma }}}=\frac{N_A}{M}\cdot {\displaystyle {\int }_V{\displaystyle {\int }_0^{\infty }m\left(\underset{\_}{x}\right)}}\cdot {\mathrm{\epsilon }}_{e_{\mathrm{\gamma }}}\left(\underset{\_}{x}\right)\cdot {\sigma }_{\mathrm{\gamma }}\left(e_n\right)\cdot \mathrm{\Phi }\left(\underset{\_}{x}.e_n\right)d{e}_{n}d\underset{\_}{x}$$

in which

• - (P _R ) _{E γ} the net count rate of the photopeak of the gamma energy element E _γ is
• M is the molar mass of the corresponding element,
• - N _A = 6.0221408 ▪ 10 ^{23 is} the Avogadro constant,
• x is the position in the sample volume V,
• - E _{n is} the neutron energy,
• m ( x ) is the distribution function for the mass of the corresponding element in the sample volume,
• - ε _{E γ} ( x ) the photopeak efficiency for gamma rays of the element at the corresponding energy E _γ emitted at position x ,
• - σ _γ (E _n ) is the partial gamma production cross section, and
• - Φ ( x , E _n ) is the neutron flux as a function of position and energy in the sample.

The partial gamma-production cross-section is dependent on the considered element. Since the irradiated elements in the sample usually emit gamma radiation at different energies, all evaluable gamma energies Eγ of an element are taken into account in the analytical evaluation for the mass determination and the uncertainty analysis on the mass determination. The consideration of several gamma energies of one element reduces the uncertainty of the measurement method.

Preferably, the sample is divided into partitions and segmented measured / analyzed. For this purpose, one or more partitions of the sample are located in the collimated field of view of the detector in a single gamma-spectrometric measurement ( 4 ). The field of view of the detector is the spatial region, which has an increased detection probability of gamma radiation due to the collimator geometry. So that the respective partitions can be aligned optimally to the field of view of the detector, the sample can be moved in front of the detector, in particular rotated and displaced. The energy-dependent detection probabilities of emitted gamma radiation from the sample or from a respective partition are referred to as photopeak efficiencies. Within the partitions of the sample, the gamma radiation is attenuated so that partitions facing the detector and located within the field of view of the collimator have higher photopeak efficiencies than partitions outside the field of view of the detector. Thus, the load capacity of a measurement result with respect to the elemental composition of a respective partition can be increased, in particular if several gamma-spectrometric measurements are taken into account.

By collimation to a partition, in particular, the SNR for each partition can be improved. If gamma-spectrometric measurements of several partitions are evaluated in combination, any remaining uncertainty can be reduced.

It has been found that special advantages arise from a number of four partitions, depending on the size of the specimen. The geometry of the partition is preferably defined as a function of the geometry of the specimen and the metrological task.

The sample is divided into partitions for cylindrical samples, in particular according to layers and sectors ( 4 ). In particular, partitions are generated in the manner of cake pieces, ie as three-dimensional cylinder segments. Horizontal partitioning (in particular, according to sections orthogonal to a symmetry axis of a cylindrical sample) is referred to as a layer, and angle-dependent partitioning is referred to as an (angular) sector. With an additional division of angular sectors as a function of the distance from the axis of symmetry, these partitions become specific as a radial sector designated. In particular, a cubic or cuboid sample can be subdivided into individual voxels. Each voxel represents a partition. The respective voxel also has a cubic or cuboid geometry.

wobei

• - ${\epsilon }_{E_{\mathrm{\gamma }}}^{i\leftarrow k}$
  
  die integrale Photopeak-Effizienz der Partition k in der Messung i ist,
• - ${\sigma }_{E_{\mathrm{\gamma }}}^{ik}$
  
  der integrale (n, γ) - Wirkungsquerschnitt für die Gamma-Energie in Partition k bei der Messung i ist, gebildet durch $${\sigma }_{E_{\mathrm{\gamma }}}^{ik}=\frac{{\displaystyle {\int }_0^{\infty }{\sigma }_{\mathrm{\gamma }}\left(E_n\right)\cdot {\mathrm{\Phi }}_k^i\left(E_n\right)d{E}_n}}{{\displaystyle {\int }_0^{\infty }{\mathrm{\Phi }}_k^i\left(E_n\right)d{E}_n}}$$
  
• - ${\mathrm{\Phi }}_k^i$
  
  der integrale Neutronenfluss in Partition k bei der Messung i ist, gebildet durch $${\mathrm{\Phi }}_k^i={\displaystyle \underset{0}{\overset{\infty }{\int }}{\mathrm{\Phi }}_k^i\left(E_n\right)d{E}_n}$$
  

The determination of the element masses in the individual partitions happens in the following way. In particular, the assumption is made that within a partition, the mass of an element mk is homogeneously distributed. With N partitions, N gamma spectra are thus recorded and N net photopeak counts result for each gamma energy. Depending on the type of partitions, variations may arise, especially in radial sectors. Then N can be replaced by K and, for example, for K partitions, N gamma spectra are recorded in N measurements, where K is greater than or equal to N, and there are N net photopeak counts for each gamma energy. Equation (1) can now, for K or N partitions, be reduced to the following sum for a gamma energy on collimated measurement of partition i, where the index K passes through the partitions, promising a simple, robust analysis: $${\left(P_R\right)}_{e_{\mathrm{\gamma }}}^i=\frac{N_A}{M}\cdot {\displaystyle \underset{k=1}{\overset{N}{\Sigma }}{m}_k}\cdot {\epsilon }_{e_{\mathrm{\gamma }}}^{i\leftarrow k}\cdot {\sigma }_{e_{\mathrm{\gamma }}}^{ik}\cdot {\mathrm{\Phi }}_k^i$$

in which

• - ${\epsilon }_{e_{\mathrm{\gamma }}}^{i\leftarrow k}$
  
  is the integral photopeak efficiency of the partition k in the measurement i,
• - ${\sigma }_{e_{\mathrm{\gamma }}}^{ik}$
  
  is the integral (n, γ) cross section for the gamma energy in partition k in the measurement i, formed by $${\sigma }_{e_{\mathrm{\gamma }}}^{ik}=\frac{{\displaystyle {\int }_0^{\infty }{\sigma }_{\mathrm{\gamma }}\left(e_n\right)\cdot {\mathrm{\Phi }}_k^i\left(e_n\right)d{e}_n}}{{\displaystyle {\int }_0^{\infty }{\mathrm{\Phi }}_k^i\left(e_n\right)d{e}_n}}$$
  
• - ${\mathrm{\Phi }}_k^i$
  
  is the integral neutron flux in partition k in the measurement i, formed by $${\mathrm{\Phi }}_k^i={\displaystyle \underset{0}{\overset{\infty }{\int }}{\mathrm{\Phi }}_k^i\left(e_n\right)d{e}_n}$$
  

wobei A eine Matrix der Dimension N×N bzw. NxK ist und m, b Vektoren der Dimension N×1 bzw. Kx1 sind. Die Einträge der Matrix A sind gegeben durch $$A_{ik}=\frac{N_A}{M}\cdot {\epsilon }_{E_{\gamma }}^{i\leftarrow k}\cdot {\sigma }_{E_{\gamma }}^{ik}\cdot {\mathrm{\Phi }}_k^i$$

Thus, equation (2) results in a system of linear equations of dimension N × N or NxK, which can be solved according to the element masses of the individual partitions. This linear equation system has the form: $$A\cdot m=b$$

where A is a matrix of dimension N × N or NxK and m, b are vectors of dimension N × 1 and Kx1, respectively. The entries of the matrix A are given by $$A_{ik}=\frac{N_A}{M}\cdot {\epsilon }_{e_{\gamma }}^{i\leftarrow k}\cdot {\sigma }_{e_{\gamma }}^{ik}\cdot {\mathrm{\Phi }}_k^i$$

Insbesondere um nur physikalisch sinnvolle, positive und eindeutige Ergebnisse bei der Lösung des Gleichungssystems zu erhalten, kann bei der numerischen Lösung des Gleichungssystems ein so genanntes „nicht negative kleinste Quadrate“-Verfahren angewendet werden.The entries of m are given by mi, and the entries of b are given by mi $$b_i={\left(P_R\right)}_{e_{\gamma }}^i,$$

In particular, in order to obtain only physically meaningful, positive and unambiguous results in the solution of the equation system, a so-called "non-negative least squares" method can be used in the numerical solution of the equation system.

wobei

• - ε_{E γ} die integrale Photopeak-Effizienz der Probe ist,
• - σ_{E γ} der integrale (n, y) -Wirkungsquerschnitt für die Gamma-Energie in der Probe ist,
• - Φ der integrale Neutronenfluss in der Probe ist.

If simplified assumptions can be made for the homogeneity of the mass distribution, the neutron flux distribution and / or the photopeak efficiencies in the entire sample for an application, in particular based on a parameter for the size and composition of the sample, a segmented / partitioned measurement is not necessarily required. It has then been found that the sample can then be measured in a single, preferably collimated measurement, and a single gamma spectrum can be recorded. Equation (2) then reduces to the following simple linear relationship: $${\left(P_R\right)}_{e_{\gamma }}=\frac{m\cdot {\epsilon }_{e_{\gamma }}\cdot N_A\cdot {\sigma }_{e_{\gamma }}\cdot \mathrm{\Phi }}{M}$$

in which

• - ε _{E γ} is the integral photopeak efficiency of the sample,
• - σ _{E γ} is the integral (n, y) cross section for the gamma energy in the sample,
• - Φ is the integral neutron flux in the sample.

This equation can be solved directly after the element mass m. The corresponding parameters can be calculated in the same way for the segmented / partitioned as well as the non-segmented case.

For each gamma energy, a mass m of the corresponding element is calculated, be it in a partition or in the entire sample volume. The uncertainty u (m) at this value is determined according to DIN ISO 11929, which can also be taken from the following publication published by the German Institute for Standardization: Determination of the characteristic limits (detection limit, detection limit and limits of the confidence interval) for measurements of ionizing radiation - basics and applications (ISO 11929: 2010) (2011).

Since an element emits gamma radiation with different gamma energies, the mass of the element is calculated as the weighted average of the individual specific masses. The weighting is based on the calculated uncertainties.

Automatic identification of the elements contained in the sample

The elements contained in the sample can be automatically identified based on the recorded gamma spectra. For each element, a characteristic emission signature of the gamma energies with corresponding intensities can be created from a nuclear physics database. The signals at the known gamma energies of the elements in the spectrum are matched by means of a computer program with this signature. A statistical analysis of the degree of correspondence provides a list of most probable elements in the sample. For this identification, not only the gamma energies of the prompt gamma radiation, but also the delayed gamma radiation are used. This results in particular the advantage that the method is applicable to a variety of elements.

Method for determining photopeak efficiencies

The determination of the energy-dependent photopeak efficiencies of an element takes place by means of the assumption of a homogeneous element and mass distribution in a partition of the sample or of the entire sample. The mean density of a partition of the sample can be determined by the mass of the sample divided by the sample volume and / or by a transmission measurement using a gamma emitter, such as Co-60 or Eu-154. The transmission measurement can serve as an extended measurement to characterize the sample, for example, to determine the filling level of a fluid to be examined or a bed in a barrel (sample) can. For the calculation of the photopeak efficiencies, points in the sample volume are randomly generated by means of a computer program, and in the next step, a large number of random trajectories are generated from this point in the direction of the detector. The path lengths in the different materials are then determined along a single trajectory and the energy-dependent probabilities of absorption and scattering along the trajectory are calculated from the attenuation coefficients of the materials. From the distance in the detector and the energy-dependent cross sections for the photoelectric reaction, the pairing and the Compton scattering becomes a Energy dependent probability calculated for complete deposition of gamma energy in the detector. Both probabilities are combined to obtain the probability of detecting a photon along the trajectory for each possible gamma energy. Averaging across all trajectories with the same starting point gives the energy-dependent photopeak efficiency for this starting point. The results of all energy-dependent photopeak efficiencies of the starting points in a partition are averaged to obtain the corresponding integral photopeak efficiency of that partition for each gamma energy.

Method for determining neutron flux and spectrum

For the analytical evaluation to determine the elemental masses of the sample, in addition to the energy-dependent photopeak efficiencies of the gamma emissions, the neutron flux and the energy-resolved neutron spectrum within the sample or the individual partitions of the sample are determined by an analytical method. In this case, a diffusion approximation of the linear Boltzmann equation is solved numerically. The input parameters for this system of equations are calculated from simulations of neutron flux and neutron spectrum in the empty sample chamber.

Traditionally, neutron flux in the sample has been determined by integrally determining energy-dependent correction factors for the entire sample. This manner of calculation has been particularly considered in the following paper: A. Trkov, G. Zerovnik, L. Snoj, and M. Ravnik: On the self-shielding factors in neutron activation analysis, Nuclear Instruments and Methods in Physics Research A , 610, pp. 553-565 (2009). For large-volume samples, an approximate method for determining the neutron flux without energy dependence was proposed in the 1994 edition of R. Overwater, detailed below, in particular for special geometries that allow a reduction to two spatial dimensions.

The neutron flux and the neutron spectrum within the sample or within individual partitions of the sample can be determined automatically (by a computer program) in the method described here, whereby a diffusion approximation of the linear Boltzmann equation is spatially and energy-resolved (considering all three spatial dimensions) numerically can be solved and the input parameters for this system of equations from simulation calculations of the neutron flux and the neutron spectrum can be calculated in the empty sample chamber.

wobei Ω die Richtungsvariable bezeichnet und Ψ somit der folgenden Gleichung genügt: $$\Omega \cdot {\nabla }_x\mathrm{\Psi }\left(\underset{\_}{x},E_n,\Omega \right)+{\mathrm{\Sigma }}_t\left(\underset{\_}{x},E_n\right)\mathrm{\Psi }\left(\underset{\_}{x},E_n,\Omega \right)={\displaystyle {\int }_{s^2}{\displaystyle {\int }_0^{\infty }{\mathrm{\Sigma }}_s\left(\underset{\_}{x},E_n^{\text{'}},E_n,\Omega \text{'},\Omega \right)}}\mathrm{\Psi }\left(\underset{\_}{x},E_n^{\text{'}},\Omega \text{'}\right)d{E}_n^{\text{'}}d\Omega \text{'}$$

The spatially and energy-resolved neutron flux in the sample can be determined from the solution of the full Boltzmann transport equation for the neutrons: $$\mathrm{\Phi }\left(\underset{\_}{x}.e_n\right)={\displaystyle {\int }_{{\mathrm{<figure-callout\; id="2"\; label="S"\; filenames="DE102017111935A1\_0027.png"\; state="\{\{state\}\}">S</figure-callout>}}^2}\mathrm{\Psi }\left(\underset{\_}{x}.e_n.\Omega \right)}d\Omega$$

where Ω denotes the direction variable and Ψ thus satisfies the following equation: $$\Omega \cdot {\nabla }_x\mathrm{\Psi }\left(\underset{\_}{x}.e_n.\Omega \right)+{\mathrm{\Sigma }}_t\left(\underset{\_}{x}.e_n\right)\mathrm{\Psi }\left(\underset{\_}{x}.e_n.\Omega \right)={\displaystyle {\int }_{s^2}{\displaystyle {\int }_0^{\infty }{\mathrm{\Sigma }}_s\left(\underset{\_}{x}.e_n^{\text{'}}.e_n.\Omega \text{'}.\Omega \right)}}\mathrm{\Psi }\left(\underset{\_}{x}.e_n^{\text{'}}.\Omega \text{'}\right)d{e}_n^{\text{'}}d\Omega \text{'}$$

Here Σt denotes the total and Σs the scattering cross section for neutrons of the sample material, which are composed of the individual cross sections of the elements contained in the sample.

Equation (6) is approximated by a coupled system of diffusion equations corresponding to the so-called SP3 approximation, which relationship was particularly considered in the following publication: PS Brantley and EW Larsen, The simplified P3 approximation, Nuclear Science and Engineering, 134, pp , 1-21 (2000). This equation system is solved numerically in multigroup form by means of a computer program. As a result, the integral neutron flux and the neutron spectrum in each partition of the sample can be obtained, resolved in the corresponding energy groups. In order to determine the input parameters, the energy resolved flow through the edge surfaces of the sample is calculated based on simulation calculations of the neutron flux in the empty measurement chamber. The parameters of the equation system result from the elemental composition of the sample, wherein in the first step the input parameters from the simulated neutron flux and the neutron spectrum in the empty sample chamber are calculated, in particular with an assumption regarding homogeneity of the sample and the elemental composition of the sample or partitions , The calculation and evaluation of the Neutron flux and spectrum can each be performed individually for each partition, in particular by defining the respective partition based on a virtual subdivision of the sample into spatial regions.

Automated iterative approach

The result of the analytical evaluation, or the most important result, is the elemental composition of the sample. The evaluation is preferably based on parameters relating to the energy-dependent photopeak efficiency, the neutron flux and the neutron spectrum within the sample or within the partitions of the sample. Based on the three input parameters shape / geometry and mass of the specimen and neutron source strength can thus take into account the three calculated parameters energy-dependent photopeak efficiencies, neutron flux and neutron spectrum within the sample and within the partitions of the sample a very comprehensive analysis. These parameters are influenced by the elemental composition of the sample, so that the process is preferably carried out iteratively until the calculated elemental composition no longer or substantially does not change significantly. With reference to 3 a possible way of iteration is described in more detail. From the input parameters, an initial neutron flux and an initial neutron spectrum are calculated (step S1 ). In addition, the exact nature of the collimation and optionally the partitioning of the sample can additionally be taken into account / determined, and furthermore also a translation and / or rotation of the sample which is advantageous for the course of the measurement can be taken into account (step S2 ). The type of collimation and the partitioning of the sample can in particular be made dependent on the size and geometry of the sample. Larger samples and complex geometries are divided into more partitions than small samples. The evaluation of the recorded gamma spectrum comprises the identification of the elements contained in the sample by an assignment of the measured peaks in the spectrum to individual elements, taking into account interferences between gamma peaks. The net count rates for the individual gamma energies are preferably calculated only once and kept constant within the iterative method (step S3 ). From this, an initial elemental composition of the sample is calculated. Using the steps described below to calculate photopeak efficiencies (step S4 ), the element masses in each partition of the sample (step S5 ) and the neutron flux and the neutron spectrum (step S6 ) determines the elemental composition of the sample. This process is iterated until the elemental composition in the individual partitions does not change. By the iterative analytical procedure, a high accuracy is achieved.

In particular, the net count rates at the individual gamma energies are calculated only once and kept constant during the iterative process. At the beginning, an initial assumption is made about the elemental composition of the sample. Using the methods described in the previous sections to calculate the photopeak efficiencies, the neutron flux, and the neutron spectrum and the element masses in each partition of the sample, a new elemental composition of the partition is then determined. This process can be iterated until the elemental composition in the individual partitions does not change.

The non-destructive method for multi-element analysis based on neutron activation described here is automatically or fully automated by this analytical procedure for the first time, especially over a longer period, especially for large-volume samples, especially already from about 1 liter. As input parameters, only the shape of the specimen and the mass are required, with the neutron source strength being taken into account in the calculation. The neutron source intensity is determined from the operating parameters of the neutron generator or from the activity of the neutron source. Also, for elemental analysis of large-volume samples, the respective sample is continuously irradiated with neutrons, and the gamma radiation emitted from the sample and the amount of an element contained in the sample are measured and evaluated after subtracting the background signal from the surface of the photopeak, the causes the element, especially in a count rate energy representation. In contrast to previous methods can be continuously irradiated and measured simultaneously, with no pulsed irradiation and no interruptions of the measuring time are necessary. It is also no longer necessary to determine the neutron flux at the location of the sample that the composition of the sample is known for at least a portion of the sample. In particular for large-volume samples, a transmission measurement for sample characterization can also be carried out.

Neutrons are to be understood as fast neutrons, which are fast when emitted from the source and which are then slowed down by a sample chamber and / or by a moderation chamber, in particular in order to be able to increase the probability of interaction with the sample. Neutrons slowed down by moderator materials can be called thermalized neutrons be designated. Thermalized neutrons are slow or decelerated free neutrons, especially with a kinetic energy of less than 100meV (milli-electron volts). In a classification for neutrons thermalized neutrons or thermal neutrons lie between the cold and the fast neutrons. The term "thermalized" indicates that the neutrons, through repeated scattering in a medium, come into equilibrium with its thermal motion. The speeds of these thermalized neutrons assume a corresponding Maxwell distribution, which can be described by a temperature.

According to one embodiment, the irradiation or the irradiation and measuring take place over a period of at least one millisecond or at least one second. In one embodiment, measuring radiation emitted in response to the irradiation during irradiation occurs in a time window of less than 5 microseconds (μsec).

According to one embodiment, the irradiation takes place over a period of at least 10 minutes, or at least 30 minutes, or at least two hours, without interruption. An optimal irradiation time must be defined depending on the specific analysis task. The continuous irradiation over such periods allows a reliable evaluation of the measurement data in a flexible manner, in particular also specifically with regard to individual sub-aspects. The irradiation time can be defined in particular with regard to a required sensitivity of the analysis task. It has been shown that the likelihood of detecting trace impurities (traces in the trace range) increases with increasing irradiation time. The sensitivity of the measuring method according to the invention can be increased with increasing time. Irradiation time and iteration can be defined independently of each other.

The minimum irradiation duration can be defined as an irradiation time in the range of seconds. The maximum irradiation duration can be defined as an irradiation time in the range of seconds, minutes, hours or even days.

According to one embodiment, neutrons are generated with the neutron energy value 2.45MeV or with at least one neutron energy value from the following group: 2.45MeV, 14.1MeV. It has been shown that with neutrons with 2.45MeV a particularly good signal can be obtained, thus a signal with favorable SNR.

According to one embodiment, neutrons are generated with a neutron energy having at least one value in the energy range 10keV to 20MeV, in particular 10keV to 10MeV.

According to one embodiment, neutrons are generated with a maximum neutron energy of 10MeV. This provides in particular a high sensitivity. As an advantage of continuous irradiation with neutron energies less than or equal to 10MeV, especially neutron energies of 2.45MeV, it has been found that many inelastic interactions do not occur as threshold responses with neutron energies of at least 3 to 4 MeV required. Such inelastic interactions degrade the SNR. By keeping the energy of the neutron source as low as possible, such inelastic interactions below the selected neutron energy can be avoided. According to one of the possible variants of the method, only the neutron energies 2.45 MeV are used.

According to one embodiment, to determine at least one element at least prompt or both prompt and delayed gamma radiation from continuous neutron irradiation is measured and evaluated. The evaluation of both types of gamma radiation extends the analysis possibilities and increases the flexibility of the method. It has been shown that it is useful to evaluate the following reaction of an irradiated atomic nucleus without having to consider a time dependence of any neutron pulse: as soon as an atomic nucleus captures a neutron, gamma radiation is automatically emitted at different energies. The atomic nucleus excites itself by emitting a cascade of gamma emissions. The generated gamma radiation is characteristic for a respective element.

According to one embodiment, to determine at least one element, only delayed gamma radiation is at least temporarily measured and evaluated in response to continuous neutron irradiation. This makes it possible to focus the research focus on certain aspects, such as a sample with a specific material.

According to one embodiment, for determining at least one element, the gamma radiation emitted from the sample is measured in an energy-resolved manner, in particular by determining photopeak count rates, wherein the determining comprises an energy-resolved evaluation of the measured gamma radiation according to at least one gamma spectrum, in particular according to a gamma spectrum detected with a respective detector. The same detector can be used for both prompt and delayed gamma radiation. The energy-resolved analysis allows flexibility and robustness. As a result, a parallel analysis of almost all elements is made possible with one measurement.

The measured gamma spectrum is specific to a detector. The respective detector may have a specific resolution each for a specific gamma spectrum.

According to one embodiment, the measuring / evaluating comprises an energy-resolved measurement / evaluation of the intensity of the gamma radiation emitted from the sample. This provides high flexibility and robustness, particularly in conjunction with evaluating multiple types of gamma radiation.

According to one embodiment, the determining comprises evaluating the measured gamma radiation, wherein the evaluating comprises correlating at least one photopeak of a count rate energy representation based on its energy with an element of the sample. This provides a comprehensive analysis of both prompt and delayed radiation, particularly for a gamma spectrum measured with one of several detectors.

In this case, the detected photopeak may be a photopeak that characterizes a prompt or delayed gamma radiation. A distinction between prompt and delayed gamma radiation in the evaluation can be made based on the respective energies, in particular also in photopeaks in which two gamma energies interfere. The determination of a respective photopeak efficiency can be carried out for both types of gamma radiation spatially and energy resolved by a numerical method in which the interactions from the location of the emission in the sample (source point) are imaged until absorption in the detector.

According to one embodiment, the evaluating comprises: quantifying the mass fraction of at least one element of the sample, in particular by evaluating the fraction of at least one element contained in the sample after subtracting a background signal from the net area of a photopeak which the element contains in a count rate energy Representation causes.

For this purpose, the photopeak can be fitted in the spectrum. The area under the photopeak can be defined as a background / background signal, and the net photo-opaque area can be defined as a useful signal.

According to one embodiment, the sample, in particular individual partitions of the sample, is / are measured collimated, in particular by means of at least one detector or by means of a plurality of detectors with a field of view collimated specifically with respect to the geometry of the partition. This can increase accuracy and also allows focusing on portions of the sample, especially with good SNR. With two or more detectors, there are advantages in particular with regard to the measuring time. It has been found that in the method described here, the measurement time can be selected the lower, the more detectors are provided, and / or that with the same measurement time, the sensitivity of the measurement can be increased.

According to one embodiment, the sample is divided into partitions and the emitted gamma radiation is measured and evaluated with respect to a respective partition. According to one embodiment, the measurement, determination and / or evaluation takes place individually with respect to individual partitions of the sample, which partitions are predefined or can be predetermined manually or automatically, in particular by collimation. This allows a focus on individual regions of the sample, or facilitates the evaluation of large-volume samples or samples with inhomogeneous composition.

Partitioning simplifies the analysis, especially with regard to a desired accuracy of the evaluation. Partitioning may, as a result, allow for spatially resolved elemental composition in the respective partitions. Partitioning also provides the advantage that assumptions can be made more easily or with less error. For example, in a cylindrical specimen eight or more partitions, in particular 12 partitions are formed, each as a cylindrical segment (cake piece). The detector unit then includes, for example, two detectors, which are not offset from each other but offset by an angle to each other (circumferential angle less than 180 °, eg circumferential angle 130 to 150 °) are arranged. The entire sample can then be fully analyzed by stepwise rotation, especially in 60 ° increments, eg in six steps using two detectors and defining 12 partitions.

It has been shown that a connection or dependency between collimation and partitioning can be used. The collimation can take place in particular as a function of the selected partitioning. A control device of the device can be set up to specify the collimation as a function of the selected partitioning. The following relationship between collimation and partitioning then be given: The entire partition is preferably in the collimated field of view of the detector. In this case, only the smallest possible proportion of space is available from other partitions, that is, from partitions that are not collimated to the target partition that is collimated, in the collimated field of view of the detector. Due to the optionally adjustable geometry of the collimator, the viewing area can be limited primarily to the target partition.

It has been shown that collimation can particularly effectively weaken the background signal. In this context, collimated measuring is to be understood as meaning, in particular, detection of the gamma radiation with at least one detector with a collimated field of vision. It has been shown that the method described here can also be carried out continuously thanks to collimation over a long period of time with particularly advantageous SNR.

According to one embodiment, the determining comprises evaluating the measured gamma radiation, wherein the evaluation is based on the assumption of a homogeneous mass and / or element distribution in the sample, in particular a homogeneous mass and / or element distribution in at least one of several partitions of the sample. This provides a robust process, especially in a highly automated iterative process.

At least in a respective partition, the element and mass distribution can be assumed to be homogeneous, so that the respective partition can be uniformly calculated / evaluated. In this case, the measurement accuracy can also be increased by geometrically selecting the partitions so that the assumption of homogeneous mass distribution applies as well as possible, e.g. no pieces of cake instead of slices in the height direction one above the other.

In contrast, previous methods often used the analytical approach that the elements are point sources. Regarding the design of the element distribution, one of two initial assumptions can be made: either point source or homogeneous element and mass distribution in a partition. It has been found that the assumption of a homogeneous element and mass distribution in connection with the method described here can lead to a very robust measurement with minimized uncertainty.

According to one embodiment, the determining comprises an evaluation of the measured gamma radiation, wherein the evaluation comprises: calculating the neutron flux within the sample, in particular within a respective partition of the sample, based on a diffusion approximation of the linear Boltzmann equation, in particular spatially and / or energy-resolved, in particular based on the following relationship: $$\mathrm{\Phi }\left(\underset{\_}{x}.e_n\right)={\displaystyle {\int }_{S^2}\mathrm{\Psi }\left(\underset{\_}{x}.e_n.\Omega \right)}d\Omega ,$$

This also provides a lean way of computation, which provides advantages, especially in iterations, in conjunction with the aforementioned advantages.

According to one embodiment, the evaluation also includes calculating the neutron spectrum within the sample, in particular within a respective partition of the sample, in particular spatially and / or energy-resolved, in particular based on the following relationship: $$\mathrm{\Phi }\left(\underset{\_}{x}.e_n\right)={\displaystyle {\int }_{S^2}\mathrm{\Psi }\left(\underset{\_}{x}.e_n.\Omega \right)}d\Omega ,$$

This provides robustness in conjunction with the aforementioned advantages.

According to one embodiment, the determining comprises evaluating the measured gamma radiation, wherein the evaluating comprises: calculating energy dependent photopeak efficiencies as well as neutron flux and neutron spectrum within the sample or within a single partition of the sample A sample, in particular, calculating neutron flux and neutron spectrum by an approximate method, each based on the following relationship: $$\mathrm{\Phi }\left(\underset{\_}{x}.e_n\right)={\displaystyle {\int }_{S^2}\mathrm{\Psi }\left(\underset{\_}{x}.e_n.\Omega \right)}d\Omega ,$$

This provides the aforementioned advantages. It has been shown that the energy-dependent photopeak efficiencies, the neutron flux and the energy-resolved neutron spectrum within the sample or within the individual partitions of the sample provide a reliable basis for the evaluation. The input parameters can be calculated from the neutron flux and the neutron spectrum in the empty sample chamber.

In particular, it has been found that the diffusion approximation enables calculation based on a small number of independent variables. This can also reduce the complexity of the analysis. A very accurate alternative method would be a numerical solution of the full linear Boltzmann equation, either deterministically or by means of a Monte Carlo method. In both variants, however, the computational effort would be very high, especially during iteration would have to be calculated with a computing time in the range of hours or even days. The diffusion approximation provides a simple mathematical structure, which allows the use of simple numerical methods.

According to an embodiment, the determining comprises evaluating the measured gamma radiation, the evaluation taking place at least partially with respect to the measured photopeak areas by detecting a plurality of photopeak areas, which are of a plurality of gamma energies respectively of at least one element in a respective partition Sample when quantifying the mass fraction of a particular element of the respective partition based on the following relationship (each for N or K partitions where the index K passes through the partitions): $${\left(P_R\right)}_{e_{\mathrm{\gamma }}}^i=\frac{N_A}{M}\cdot {\displaystyle \underset{k=1}{\overset{N}{\Sigma }}{m}_k}\cdot {\epsilon }_{e_{\mathrm{\gamma }}}^{i\leftarrow k}\cdot {\sigma }_{e_{\mathrm{\gamma }}}^{ik}\cdot {\mathrm{\Phi }}_k^i$$

In other words, in the evaluation, a plurality of gamma energies of at least one element in each partition of the sample can be analyzed in quantifying the mass fraction of each element. This provides a slim, robust, flexible process with good accuracy. It can ensure a high quality of the measurement / evaluation.

The neutron flux in the sample has so far been determined by determining energy-dependent correction factors integrally for the entire sample. It has been shown, in particular for large-volume samples, that an approximate method for determining the neutron flux without energy dependence and for specific geometries that allow a reduction to two spatial dimensions can also be used. Such a method is based in particular on a diffusion equation which is determined by only two parameters. In particular, aspects of a method can be applied, the relationships of which have already been considered in detail in the following publication: R. Overwater, The Physics of Big Sample Instrumental Neutron Activation Analysis, Dissertation, Delft University of Technology, Delft University Press, ISBN 90-407- 1048-1 (1994).

In contrast, the neutron flux and the neutron spectrum within the sample or individual partitions of the sample can be determined by a computer program according to the present method, whereby a diffusion approximation of the linear Boltzmann equation is solved numerically and spatially resolved numerically, especially considering all three spatial dimensions. The boundary conditions for this system of equations can be calculated from simulation calculations of the neutron flux in the empty sample chamber. Energy-dependent correction factors are not required or need not be defined.

The calculation and evaluation of the neutron flux and spectrum can each take place individually for a respective partition, in particular by defining the respective partition based on a virtual subdivision of the sample into spatial regions.

According to one embodiment, the method is carried out based on the input parameters neutron source strength, sample geometry and sample mass, in particular exclusively based on these three input parameters. This can automate the process to a high degree. There are then only three input parameters to specify. Further parameters can be determined numerically / automatically. As a result, the effort on the part of a user can be minimized. Further input parameters can be provided, for example, by nuclear physics data or by the simulative calculation of the input parameters for the calculation of neutron flux and neutron spectrum.

Optionally, a monitor, or a calibration material of prior art composition, may be analyzed along with the sample. This can increase the measurement accuracy or the load capacity of the measurement, in particular for samples whose composition is unknown, or for samples in which assumptions are made which are rather uncertain. However, the use of a monitor and the evaluation of gamma radiation emitted by the monitor may optionally be done independently of the aspects of the method and apparatus described herein. In particular, a material can be used which is certainly not found in the sample, for example, gold in the form of a very thin gold foil, which is placed on the sample.

According to one embodiment, the method is carried out automatically, in particular by evaluating the measured gamma radiation based on apart from the three parameters neutron source intensity in the irradiation, sample geometry and sample mass, exclusively numerically determined parameters. This provides autarky and the ability to easily iterate. The process becomes more robust.

Also with regard to the three aforementioned parameters can optionally be an automation. The sample geometry can be detected autonomously by a camera unit, and the sample mass by a weighing unit. Both components camera unit and weighing unit are preferably not in the interior of the device, so not within the neutron field, but outside the sample chamber, in particular positioned outside of the shield. For this purpose, the device may have a measuring station for sample specification, at which measuring station the sample can be automatically characterized. The neutron source intensity can be obtained directly as a controlled variable from the neutron generator. The neutron source strength is directly dependent on the high voltage and the current intensity of the neutron generator. Through this expansion of automation, a very self-sufficient and user-friendly device or measuring system can be provided.

According to one embodiment, at least one measurement from the following group is carried out to characterize the sample: transmission measurement, sample weighing, optical detection of the sample geometry. On the one hand this facilitates the handling of the measuring method for the user, and on the other hand it also facilitates the subsequent measurement, in particular also with regard to a partitioning.

According to one embodiment, the method is carried out iteratively, in particular in each case with respect to individual elements or with respect to the complete composition of the sample or with regard to individual partitions of the sample. This provides good accuracy, especially in an easy-to-use process. This also provides a method with a high degree of autarchy.

At least one of the aforementioned objects is also achieved by a method of multi-element analysis based on neutron activation, comprising the steps of: generating fast neutrons with energy in the range of 10keV to 10MeV; Irradiating a sample with the neutrons;

Measuring the gamma radiation emitted from the irradiated sample to determine at least one element of the sample; wherein the irradiation of the sample is carried out unpulsed in a continuous manner, wherein the measuring takes place independently of the irradiation during the irradiation, in particular simultaneously for irradiation, wherein at least prompt or both prompt and delayed gamma irradiation from continuous neutron irradiation is measured and evaluated to determine at least one element wherein the evaluation is based on the assumption of a homogeneous mass and / or element distribution in the sample or in at least one of a plurality of partitions of the sample. This results in many of the aforementioned advantages. The measurement / evaluation can take place independently of the time course of the irradiation or independently of individual phases of an irradiation.

At least one of the aforementioned objects is also achieved by using a detector unit having at least one detector in the multi-element analysis of a sample based on neutron activation configured to continuously measure both prompt and delayed gamma radiation due to continuous irradiation of the sample with neutrons, the gamma radiation at least partially also continuously, irrespective of the time of irradiation and independent of any Neutron pulses, in particular without time window, and simultaneously measured for continuous irradiation, in particular use of the detector unit with a plurality of detectors respectively collimated or partition collimated or adjustable collimatable with respect to at least one partition or at least one predefinable geometry of a partition, preferably with a collimator of lead or bismuth. This results in the aforementioned advantages.

At least one of the aforementioned objects is also achieved by using at least one neutron source for multi-element analysis of a sample based on neutron activation for generating fast neutrons for continuously irradiating the sample with first neutrons having at least one neutron energy value from the following group: 2.45MeV, 14.1MeV; and / or with second neutrons having a neutron energy with at least one value in the energy range 10 keV to 20 MeV, in particular 10 keV to 10 MeV; and / or with third neutrons with a maximum neutron energy of 10MeV. This results in the aforementioned advantages. Preferably, the emitted gamma radiation is detected with a detector with a collimator of lead or bismuth.

At least one of the aforementioned objects is also achieved by a control device configured to drive at least one neutron generator of a device for multi-element analysis based on neutron activation, in particular a device described here, wherein the neutron generator is adapted to generate fast neutrons with energy in the range of 10keV to 20MeV , in particular 10keV to 10MeV, wherein the control means is arranged to drive the neutron generator to generate the neutrons and to irradiate the sample in an unpulsed continuous manner, in particular during at least a first time window, and wherein the control means is further adapted to drive at least one detector to the continuous one and / or temporarily measuring gamma radiation emitted from the sample or a single partition of the sample simultaneously with the irradiation, in particular during at least a second time itfensters independent of the first time window, continuously simultaneously to the continuous irradiation and / or time-independent thereof. The first and second time windows can be different or can be predefined or set independently of one another. The control device is in particular configured to control a previously described method.

This type of analysis provides previously mentioned benefits. The control device can synchronize at least the irradiation, measuring and optionally also the positioning of the sample (in particular by activating / controlling a rotary / lifting device) and thus control the actual measuring method of the measuring system, in particular the way and the period of data collection. By means of this control device, the irradiation and measuring, in particular over a period of e.g. at least 20 or 50sec. take place, or even over several hours or days. The control device can be coupled to a rotary / lifting device and further be set up for positioning a sample arranged on a sample carrier by means of the rotary / lifting device, in particular according to or in dependence on the geometry of partitions of the sample. This not least provides the possibility to control by means of a single control device, the entire device or to run the entire process (three-function control device) comprising driving the neutron generator, driving the at least one detector, and positioning of the sample. The neutron generator may be e.g. Parameters such as the neutron source strength are given, and the rotary / lifting device position data or displacement paths and displacement speeds can be specified.

The controller may be configured to control the operation of a neutron generator configured to merge deuterons to generate fast neutrons for multi-element analysis of a sample by continuous, unpulsed irradiation of the sample.

The device or controller may include an input mask (user interface) or input unit for manually inputting the following three parameters: neutron source intensity in the irradiation, sample geometry, and sample mass. These parameters can also be stored in a data memory and read out by the control device and transferred to a computer program product.

At least one of the aforementioned objects is also achieved by a computer program product for multi-element analysis based on neutron activation, and arranged to determine at least one element of an unpulsed continuously neutrally irradiated sample by evaluating emitted from the sample gamma radiation, namely prompt and / or delayed gamma radiation, based on energy-dependent photopeak efficiencies as well as neutron flux and neutron spectrum within the sample or within a single partition of the sample, in particular configured for partition collimated evaluation of measured gamma radiation based on the following relationship: $${\left(P_R\right)}_{e_{\mathrm{\gamma }}}^i=\frac{N_A}{M}\cdot {\displaystyle \underset{k=1}{\overset{N}{\Sigma }}{m}_k}\cdot {\epsilon }_{e_{\mathrm{\gamma }}}^{i\leftarrow k}\cdot {\sigma }_{e_{\mathrm{\gamma }}}^{ik}\cdot {\mathrm{\Phi }}_k^i$$

This formula relates to the determination of element masses according to the invention in a single partition. This computer program product allows a high degree of automation or autarky, in conjunction with a high accuracy of the analysis. The computer program product is in particular set up for automatically carrying out a previously described type of evaluation. As partition collimated evaluation, an evaluation is to be understood individually with regard to individual partitions of the sample, which partitions were previously geometrically defined and delimited from each other.

In addition, the computer program product can also be set up for specifying a desired position of the sample, in particular as a function of a detected or entered sample geometry, in particular based on setpoint positions stored in a position database as a function of the sample geometry and / or sample size, the positioning being determined in particular by activation / Rules of a rotary / lifting device can be done.

The computer program product can be set up to calculate different variants of a partitioning for a respective sample geometry and to propose a partitioning identified as optimally or to select it directly autonomously. For this purpose, the computer program product can carry out a rough uncertainty analysis as a function of the number and configuration of the partitions and determine or autonomously determine the partitioning as a function of the accuracy required by the user. With this procedure, a configurable collimator can be set specifically with regard to the selected partitions.

Accordingly, at least one of the aforementioned objects is also solved by a data carrier having such a computer program product stored thereon, or by a computer or a computer system or a virtual machine or at least one hardware element with it.

Accordingly, at least one of the aforementioned objects is also achieved by a computer program designed to provide the manner and manner of evaluation described herein or the respective method steps described herein.

According to one embodiment, the computer program product is configured to evaluate an integral measurement, in particular with respect to a non-partitioned sample based on a single gamma spectrum, in particular based on the following relationship: $${\left(P_R\right)}_{e_{\mathrm{\gamma }}}=\frac{m\cdot {\epsilon }_{e_{\gamma }}\cdot N_A\cdot {\sigma }_{e_{\gamma }}\cdot \mathrm{\Phi }}{M}$$

This provides a simplification regarding the assumption of a homogeneous mass distribution.

The process according to the invention can also be described as follows. In a method for non-destructive elemental analysis of samples, the respective sample is continuously irradiated with fast neutrons, wherein simultaneously to the irradiation, the emitted / emitted by the sample gamma radiation is measured, wherein the amount of an element contained in the sample after deduction of the background signal from the net area of Photopeaks is evaluated, which causes the element in a count rate energy representation. In this case, the quantification of the elemental masses of a sample can be automated, and the parameters required for the analysis, except for the neutron source intensity in the irradiation and the geometry and total mass of the sample, can be calculated numerically. A monitor for the neutron flux or a sample internal or external calibration standard is not required. The sample can be decomposed into space regions (partitions) and each partition of the sample can be measured collimated. The determination of the element masses of the sample can be based on the fact that the elemental and mass distribution in individual partitions of the sample is assumed to be homogeneous. The average density of a partition can be measured by a transmission measurement with a radioactive gamma emitter. In this case, the neutron flux within the partitions of the sample can be determined by an analytical method, which is a Diffusion approximation of the linear Boltzmann equation solves numerically and calculates boundary conditions for this system of equations from simulation calculations of the neutron flux in the empty sample chamber. The determination of photopeak efficiencies can be spatially and energy resolved by a numerical method in which the interactions of the emission in the sample (source point) to the absorption in the detector are mapped. In doing so, the method can be carried out iteratively with respect to the composition of the sample until the calculated composition of the sample stabilizes. In this case, all detected photopeak areas which are produced by different gamma emissions of an element in the sample can be taken into account in the analytical evaluation. In doing so, the results of the measurement of each partition can be taken into account in the analytical evaluation, whereby the sensitivity and the accuracy of the measurement procedure for the entire sample can be improved. The neutron source and the sample may be located in a sample chamber made of graphite as a moderation chamber. In this case, an effective shield for neutron radiation can be located directly around a moderation chamber or around the sample chamber. In this case, the detector or the detector unit may be located in a collimator of gamma-rays shielding material. The geometry of the sample and moderation chamber and / or of the sample carrier can reduce the neutron flux gradient within the sample, which can be achieved in particular by means of variable moderation lengths (distance between neutron source / neutron source point and sample), with the effect that the neutron flux gradient is reduced or changed becomes.

• einem Neutronengenerator eingerichtet zum Generieren von schnellen Neutronen;
• einer Probenkammer und einem darin angeordneten Probenträger;
• einer Detektoreinheit mit wenigstens einem Detektor eingerichtet zum Messen von aus einer bestrahlten Probe emittierter Gammastrahlung zum Bestimmen wenigstens eines Elements der Probe; wobei die Vorrichtung eingerichtet ist zum ungepulsten kontinuierlichen Bestrahlen einer/der auf dem Probenträger angeordneten Probe und eingerichtet ist zum Messen von aus der bestrahlten Probe emittierter prompter und/oder verzögerter Gammastrahlung zeitunabhängig von der Bestrahlung, insbesondere ohne Zeitfenster, während des Bestrahlens, insbesondere simultan zum kontinuierlichen Bestrahlen, insbesondere eingerichtet zum Ausführen eines zuvor beschriebenen Verfahrens. Hierdurch ergeben sich zuvor genannte Vorteile, insbesondere in Verbindung mit einem niedrigen oder vorrichtungs- und/oder verfahrenstechnisch minimierten Untergrundsignal.

At least one of the aforementioned objects is also achieved by a device for multi-element analysis based on neutron activation, comprising:

• a neutron generator adapted to generate fast neutrons;
• a sample chamber and a sample carrier disposed therein;
• a detector unit having at least one detector configured to measure gamma radiation emitted from an irradiated sample to determine at least one element of the sample; wherein the device is arranged for unpulsed continuous irradiation of a sample arranged on the sample carrier and is arranged for measuring prompt and / or delayed gamma radiation emitted from the irradiated sample, irrespective of the time of irradiation, in particular without time window, during the irradiation, in particular simultaneously continuous irradiation, in particular adapted for carrying out a method described above. This results in the aforementioned advantages, in particular in conjunction with a low or device and / or procedurally minimized background signal.

As a single detector (a detector unit), a semiconductor or scintillation detector is preferably used, ie a detector with high energy resolution, which is set up for the measurement of the prompt and delayed gamma radiation.

Optionally, the method is varied by means of a moderation chamber provided independently of the sample chamber. By default, the moderation chamber can be provided / installed in the device in a stationary manner. The process of moderation can be performed in the desired manner in the sample chamber, in the moderation chamber and / or in the sample itself.

The respective detectors of the device can be focused by at least one collimator. A collimator adapted to predetermining or narrowing or adjusting the field of view of the detector provides in particular the advantage of an improved SNR, especially in connection with continuous irradiation. In addition, the sample can be measured partition collimated.

Preferably, a plurality of detectors are arranged in the same height level, in particular in the height level of the neutron source or the neutron source point. Preferably, the detector or detectors are arranged as close as possible to the neutron source point. This provides good measurement results or enables the minimization of background signals. In a plan view with an arrangement in the same height level, the detectors are preferably offset by less than 90 ° in the circumferential direction relative to the neutron source point, for example by 60 or 75 °.

The neutron source point here is preferably the location or the position at which the neutrons are emitted, in particular emitted into the sample chamber onto the sample. The neutron generator can be arranged independently of the position of the neutron source point, or specify the position of the neutron source point.

The collimator, as well as a moderation chamber, can be permanently installed in a single predefined setting or configuration. Optionally, the collimator may also have multiple adjustments, each for a predefinable field of view, eg, a first setting with a relatively wide / wide field of view, a second setting with a medium field of view, and a third setting with a relatively narrow / narrow / focused field of view the collimator can be switched between the settings.

According to one exemplary embodiment, the device comprises at least one component weakening a background signal of the device from the following group: at least one collimator of lead or bismuth, preferably a collimator of the (respective) detector on the sample, in particular on a partition of the sample or a graphite moderation chamber, and / or a screen of borated polyethylene, and / or a sample chamber and / or a sample carrier in each case at least partially of graphite or completely fluorinated plastics or beryllium. In this way, in particular, an improved SNR can be achieved. The collimator preferably has a wall thickness of at least 5 cm. The device can analyze the sample nondestructively and evaluate the emitted gamma radiation in many aspects. The device is not limited to the evaluation of a particular type of gamma radiation or in a specific time window.

It has been shown that an advantageous angle between neutron generator and detector is between 50 and 90 °, in particular since this can prevent the detector from being exposed to too high a neutron flux. In addition, the detector can be focused in this angular range to a spatial area in which the neutron flux in the sample is maximally high.

Previously, gamma radiation from inelastic interactions (scattering processes) in previously used systems had produced such a strong background signal that the detection of gamma radiation was only possible after a certain waiting time (time window) after a neutron pulse. Previously used detectors and the detectors in previously used systems were so far almost "short" after a respective neutron pulse. The too high signal rate caused the detector to be paralyzed. The detector then required a certain waiting time until a detection can take place again, ie until the signal rate has fallen again.

According to one embodiment, the apparatus further comprises a computer program product or data store therewith, wherein the computer program product is adapted to determine at least one element of the sample by evaluating the measured gamma radiation based on energy dependent photopeak efficiencies, neutron flux and neutron spectrum within the sample or within a single partition of the sample Sample, in particular based on at least one of the previously described with respect to the computer program product relationships. These formulas or relationships can each be stored as one of several calculation bases in a data memory, by means of which the computer program product interacts or accesses. In addition to high flexibility in the determination / evaluation, this also provides the option of a fully automatic, iterative multi-element analysis, in particular in conjunction with a control device for controlling / regulating neutron emission, detectors and / or a rotary / lifting device.

According to one embodiment, the device further comprises a rotating and / or lifting device arranged for translational and / or rotational displacement of the sample carrier or the sample, preferably a decoupled from a / of the sample chamber of the device rotating and / or lifting device, wherein at least one electrical Drive the rotating / lifting device outside of a shield (in particular outside a borated polyethylene screen) of the device is arranged. This also provides a good SNR.

According to one embodiment, the device further comprises a unit for measuring transmission, which is set up for determining the average density of the sample or the respective partition. The transmission unit comprises a radioactive gamma emitter, in particular Eu-154 or Co-60, and a detector for measuring the attenuation of the gamma radiation after penetration of the sample. The detector for measuring the attenuation of the gamma radiation may be one of the detectors for the prompt and delayed gamma radiation, or may also be a detector provided specifically for this transmission measurement. During the transmission measurement, the sample is not irradiated with neutrons. This makes it possible to distinguish the type of gamma radiation.

According to one embodiment, the device comprises at least two detectors, in particular in a symmetrical arrangement relative to the neutron generator and / or relative to at least one neutron source or to at least one neutron source point of the device. As a result, the respective partition can be optimally positioned or aligned in front of a respective detector. The partitions can also be defined geometrically as a function of the geometry of the sample, and the sample can be aligned accordingly, for example, rotated once in full in six rotation steps.

The translation and / or rotation may be performed such that the center of each partition is at the level of the detector or on the viewing axis of the detector, or such that the sample or partition is in a collimated viewing area of the detector.

According to one embodiment, the device further comprises a control device configured for automated continuous irradiation and / or set up for controlling automatic measurement with continuously applied neutron irradiation time-independent of the neutron irradiation simultaneously to the continuous irradiation, in particular adapted for iterative automated evaluation of the emitted and measured gamma radiation time independent of the neutron irradiation based on apart from the three manually or automatically specifiable parameters neutron source strength in the irradiation, sample geometry and sample mass, exclusively numerically determined or read by the controller parameters. This provides the opportunity to run individual steps or even the entire process automated.

According to one exemplary embodiment, the neutron generator comprises at least one neutron source or a neutron source point configured for the fusion of deuterons (deuterium nuclei), in particular with deuterium gas as gaseous target or gaseous operating substance. It has been shown that a sufficiently high swelling strength can also be ensured by the fusion of deuterons. The use of this energy range provides advantages in continuous irradiation and also in measuring and evaluating, on the one hand due to low neutron energy, on the other hand with regard to a long irradiation time. The fuel may be gaseous (instead of solid), so that it is no longer necessary to replace a solid target (solid) consumed after a certain period of time. This particularly simplifies the analysis over a long period of time, and can ensure high reproducibility or high reliability of the measurement results.

According to one embodiment, the neutron generator is an electrically operated neutron generator or comprises at least one radionuclide neutron source, such as e.g. an AmBe source. Preference is given to a neutron generator which fuses deuterons and emits neutrons with a starting energy of 2.45 MeV by means of this fusion reaction. Optionally, a pulsed irradiation can be carried out especially at 2.45MeV. In contrast, with previous pulsed irradiation, a neutron generator is often used to fuse tritium-deuterium (14.1 MeV), for neutrons with exactly this one energy value.

According to one embodiment, the at least one detector is a semiconductor or scintillation detector. This allows accurate evaluation of both prompt and delayed gamma radiation in a wide range of energies.

• einem Neutronengenerator eingerichtet zum Generieren von schnellen Neutronen;
• einer Probenkammer und einem darin angeordneten Probenträger;
• einer Detektoreinheit mit wenigstens einem Detektor eingerichtet zum Messen von aus einer bestrahlten Probe emittierter Gammastrahlung zum Bestimmen wenigstens eines Elements der Probe; wobei die Vorrichtung eingerichtet ist zum ungepulsten kontinuierlichen Bestrahlen einer/der Probe und eingerichtet ist zum Messen von aus der bestrahlten Probe emittierter prompter und/oder verzögerter Gammastrahlung zeitunabhängig von der Bestrahlung während des Bestrahlens, wobei die Vorrichtung wenigstens eine ein Untergrundsignal der Vorrichtung schwächende Komponente aus der folgenden Gruppe umfasst: wenigstens einen den Sichtbereich des Detektors auf die Probe bzw. Partition einschränkenden Kollimator, und/oder eine Moderationskammer aus Graphit, und/oder eine Abschirmung aus boriertem Polyethylen, und/oder eine Probenkammer und/oder einen Probenträger jeweils zumindest teilweise aus Graphit oder vollständig fluorierten Kunststoffen oder Beryllium, wobei die Vorrichtung ferner umfasst eine Steuerungseinrichtung eingerichtet zum automatisierten kontinuierlichen Bestrahlen und/oder eingerichtet zum Steuern/Regeln eines automatisierten Messens bei kontinuierlich applizierter Neutronenbestrahlung zeitunabhängig von einzelnen Phasen der Neutronenbestrahlung während des Bestrahlens, insbesondere simultan zum kontinuierlichen Bestrahlen. Hierdurch ergeben sich zahlreiche der zuvor genannten Vorteile.

At least one of the aforementioned objects is also achieved by a device for multi-element analysis based on neutron activation, comprising:

• a neutron generator adapted to generate fast neutrons;
• a sample chamber and a sample carrier disposed therein;
• a detector unit having at least one detector configured to measure gamma radiation emitted from an irradiated sample to determine at least one element of the sample; wherein the apparatus is arranged for unpulsed continuous irradiation of a sample and is adapted to measure prompt and / or delayed gamma radiation emitted from the irradiated sample, independently of the irradiation during irradiation, the apparatus comprising at least one sub-component signal attenuating component of the following group comprises at least one collimator restricting the field of view of the detector to the sample or partition, and / or a graphite moderation chamber, and / or a screen of borated polyethylene, and / or a sample chamber and / or a sample carrier in each case at least partially graphite or fully fluorinated plastics or beryllium, the device further comprising a Control device configured for automated continuous irradiation and / or set up to control / regulate an automated measurement with continuously applied neutron irradiation time-independent of individual phases of the neutron irradiation during irradiation, in particular simultaneously for continuous irradiation. This results in many of the aforementioned advantages.

By using a neutron generator adapted for the fusion of deuterons, in particular with deuterium gas as a gaseous target or gaseous fuel, for generating fast neutrons for multi-element analysis of a sample based on neutron activation, in particular for continuous, unpulsed irradiation of the sample, the aforementioned advantages can be realized ,

list of figures

• 1 in einer perspektivischen Ansicht in schematischer Darstellung eine Vorrichtung zur zerstörungsfreien Multielementanalyse gemäß einem Ausführungsbeispiel;
• 2A, 2B, 2C jeweils in einer Schnittansicht eine Probenkammer mit einem bzw. zwei Detektoren sowie in Detailansicht einen Detektor jeweils einer Vorrichtung zur zerstörungsfreien Multielementanalyse gemäß einem Ausführungsbeispiel;
• 3 in schematischer Darstellung im Flussdiagramm einzelne Schritte eines Verfahrens gemäß einer Ausführungsform;
• 4 eine zylindrische Probe mit einer Partitionierung in Form von Scheibensegmenten, als ein Beispiel für eine Partitionierung bei einem Verfahren gemäß einer Ausführungsform; und
• 5 in einer Schnittansicht eine Probenkammer mit einer außerhalb einer Abschirmung angeordneten Dreh- und Hubeinrichtung einer Vorrichtung gemäß einem Ausführungsbeispiel.

In the following drawing figures, the invention will be described in more detail, reference being made to reference characters, which are not explicitly described in a respective drawing figure, to the other drawing figures. Show it:

• 1 in a perspective view in schematic representation of a device for non-destructive multi-element analysis according to an embodiment;
• 2A . 2 B . 2C in each case in a sectional view a sample chamber with one or two detectors as well as in detail a detector of a device for non-destructive multi-element analysis according to an embodiment;
• 3 in a schematic representation in the flowchart individual steps of a method according to an embodiment;
• 4 a cylindrical sample with partitioning in the form of disk segments, as an example of partitioning in a method according to an embodiment; and
• 5 in a sectional view of a sample chamber with a arranged outside a shield rotation and lifting device of a device according to an embodiment.

DETAILED DESCRIPTION OF THE FIGURES

In the 1 is an assembly of a device 10 for non-destructive multi-element analysis, specifically in the manner of a measuring system for carrying out the method described here for multi-element analysis based on neutron activation.

By the operation of one or more neutron generators 11 becomes a sample 1 irradiated continuously with neutrons and simultaneously measured for irradiation induced thereby / emitted gamma radiation. The device / measuring system 10 including sample 1 consists in particular of the following assemblies: The neutron generator 11 comprises at least one electrically operated neutron source, in particular a neutron source, which at least fuses deuterium and deuterium (or deuterons), and optionally allows a further type of fusion, in particular tritium and deuterium. The deuteron fusion reaction emits fast neutrons with an energy of 2.45MeV. Deuterium gas is preferably used as target (non-radioactive). Optionally, at least one additional energy value, in particular 14.1 MeV, can also be provided by means of the neutron generator. The neutron generator 11 is inside a moderation chamber 12 and is surrounded by a shield 19 , The moderation chamber 12 consists of a material which effectively moderates fast neutrons and which emits as little gamma radiation as possible during a moderation process, preferably graphite. Gamma radiation that is not emitted from the sample and yet registered by the detector is defined as an active background signal. The device described here 10 advantageously provides a very weak, minimized background signal so that gamma radiation can be measured in a very flexible manner.

The sample 1 is located on a sample carrier during the irradiation 14 inside a sample chamber 15 , The sample carrier may be, for example, a turntable, a box, a can or a bottle. Preferably comes as a material for the sample carrier 14 Graphite and fully fluorinated plastic used.

The sample carrier 14 and the sample chamber 15 are designed so that the neutrons irradiate the sample as homogeneously as possible (ie with a low local neutron flux gradient) and that neutrons that escape from the sample are effectively reflected back into the sample. In the interactions between the neutrons and the sample carrier 14 as well as the sample chamber 15 should one possible only weak active background signal. This can be ensured in particular by using as material for the sample carrier 14 and the sample chamber 15 Preferably, graphite, beryllium and fully fluorinated or carbon fiber reinforced plastics are used.

The gamma spectrum measured simultaneously for the irradiation is detected by a detector unit 16 or by one or more detectors 16A . 16B added. As a detector unit 16 Both one and more detectors can be understood. Several detectors can be used to reduce the measuring time of a sample or, with the same measuring time, the sensitivity and accuracy of the multi-element analysis method can be increased. The detector unit 16 registers the energy of the gamma radiation emitted from the sample and counts the energy depositions in the detector. To a respective detector 16 There is a collimator around 17 , By means of the respective collimator, the "field of view" of the detectors used can be restricted in such a way that mainly gamma radiation emitted from the sample is detected. The spatial area with increased detection probability for gamma radiation has, starting from the detector, in particular the shape of a cone or a pyramid. The collimator 17 is made of a material which shields gamma radiation as effectively as possible, preferably lead. Thanks to the limited field of view of the detector, the collimator makes it possible to minimize or weaken the active background signal.

The sample can be measured segmented / partitioned. For this purpose, in a single gamma spectrometric measurement, not the entire sample body but only individual segments, so-called partitions, are located in the collimated field of view of the detector. To position the individual partitions in the field of view of the detector for rotation and / or translation of sample and sample carrier a rotating and lifting device 18 intended. The rotary and lifting device and the sample carrier are connected to each other in particular by force and / or positive locking. Since the components of the rotary and lifting device could increase the active background signal, these assemblies are preferably outside the sample and moderation chamber 15 . 12 and outside the shield 19 positioned ( 5 ). For power transmission between the rotary and lifting device and the sample carrier 14 In particular, a shaft, chain or a toothed belt can be used.

The shield 19 is around the moderation and sample chamber 12 . 15 as well as the detector unit 16 and the collimator 17 arranged around. The shield 19 encloses the measuring system and reduces the neutron and gamma local dose rate outside the measuring system. Preferably comes as a material for the part of the shield, which shields primarily neutron radiation, borated polyethylene used. As materials for the part of the shield which primarily reduces or attenuates gamma radiation, concrete and high atomic number and high density elements such as steel or lead may be used. It has been shown that borated polyethylene in the moderation and sample chamber 12 . 15 or around it, as well as around the detector unit 16 and the collimator 17 the SNR can be significantly improved.

In 1 is further indicated that at the input mask 23 at least one of at least three variables / parameters v1 . v2 . v3 can be entered or retrieved, in particular the neutron source strength, the sample geometry, and / or the sample mass. These three parameters can according to a variant also fully automatically by the device 10 be determined.

At the in 1 shown arrangement may be a moderation in the separate moderation chamber 12 outside the sample chamber 15 be performed. Optionally, even within the sample chamber 15 be moderated. A moderation can generally in the moderation chamber 12 in the sample chamber 15 and / or in the sample 1 yourself.

In 1 is also a transmission measuring unit 24 shown, by means of which optionally an additional characterization of a sample can be carried out, in particular based on gamma irradiation.

In 1 Furthermore, components for automating the measurement or evaluation are indicated, in particular a control device 20 which is coupled to a data memory 21 , to a nuclear physics database 22 to an input unit / mask 23 to a transmission measuring unit 24 , to a camera unit 25 to a weighing unit 27 , and / or a computer program product 30 the latter also in the control device 20 can be deposited.

2A . 2 B . 2C show a device for non-destructive multi-element analysis, by means of which a collimated measurement of partitions of a sample takes place. At the in 2A shown variant are the two detectors 16A . 16B symmetric relative to a neutron source or to a neutron source point 11.1 of the neutron generator 11 arranged.

2A . 2 B . 2C show in detail the preferred materials, in particular materials for ensuring a weak background signal, in particular borated polyethylene M1 (in particular 5 or 10%) for the shielding 19 of neutron radiation (sometimes also concrete for the shielding 19 of gamma radiation), lead or bismuth M2 for the collimator 17 or for the purpose of shielding gamma radiation, graphite M3 for the moderation chamber 12 or the sample carrier 14 or the sample chamber 15 , Lithium 6 Polyethylene or lithium 6-silicone M6 for shielding the detector, germanium M10 for the crystal 16.1 , An area between individual components of the detector 16 , in particular between a detector end cap 16.2 and the crystal 16.1 , is with air M4 filled, in particular inside the collimator. Depending on the nature of the neutron generator or detector, adequate materials may be available for individual further components M7 . M8 are selected, in particular from the list comprising copper, aluminum, plastic.

2A shows an example of two partitions P1 . P2 of n partitions pn in the form of cylinder segments (cake pieces), in a cylindrical sample 1 , The sample 1 can be provided by a barrel, for example, with a certain filling height of a bed or a fluid. The barrel can have a fairly large volume, eg 200 liters.

In 2A is also the orientation of the individual components according to the longitudinal axis x , the transverse axis y and the vertical axis or vertical axis z to recognize. The sample is at least partially cylindrical or, for example, designed as a barrel and extends along the vertical axis z , in particular rotationally symmetrical about the z-axis. Positioning in different z-height positions is by means of the aforementioned lifting device 18 possible.

2 B shows a variant with only one detector 16 , which is collimated on a cylinder segment. This arrangement can also be provided in a cost-optimized manner.

At each in 2A . 2 B arrangement shown, a moderation also only within the sample chamber 15 be performed.

2C further shows a detector end cap 16.2 and a crystal holder 16.3 by which elements the crystal 16.1 can be positioned and aligned.

3 shows a method in six steps S1 to S6 which steps each comprise sub-steps. Between the individual steps can control points R1 to R5 be it for a user query, be it for an automatic, computer-controlled query.

In a first step S1 there is a generation of neutrons and irradiation of a sample with neutrons, wherein the first step may comprise at least one of the following substeps: adjusting (controlling or regulating) the neutron source intensity ( S1.1 ), Moderation ( S1.2 ), simulative calculation of the neutron spectrum ( S1.3 ), simulatively calculating the neutron flux ( S1.4 ). In a first regulation point R1 In particular, an optionally repeated query can be made with respect to the neutron source intensity, be it an automated data query, be it in the context of user input / user guidance.

In a second step S2 there is a sample specification and measurement, wherein the second step may comprise at least one of the following substeps: detecting the sample mass and optionally also the sample geometry ( S2.1 ), Collimation ( S2.2 ), Detecting or adjusting the sample partitioning ( S2.3 ), Moving / positioning the sample, in particular by translation and / or rotation ( S2.4 ). The step S2.1 may be in conjunction with a transmission measurement, in particular by emitting radioactive gamma radiation to the sample, for example to detect a fill level in a drum (sample), or to determine a matrix density. The transmission measurement can therefore be understood as an extended measurement for characterizing the sample, and can provide further data, in particular also with regard to partitioning which is as expedient as possible. In a second regulation point R2 In particular, an optionally repeated query regarding the sample mass, sample geometry and partitioning can be made, be it an automated data query in communication with a camera unit and / or a weighing unit, be it in the context of user input / user guidance. In the second control point R2 In particular, a positioning or alignment of the sample can take place.

In a third step S3 there is a detection or measurement of emitted gamma radiation, wherein the third step may comprise at least one of the following substeps: detection / measurement of gamma radiation and evaluation of the gamma spectrum ( S3.1 ), Element / peak identification ( S3.2 ), Interference analysis ( S3.3 ), Evaluation of peaks, in particular evaluation in terms of area and background ( S3.4 ). In a third regulation point R3 In particular, a transfer and review of interim results can take place. This can be the control point R3 include a plausibility check, in particular in the context of a statement about the homogeneous element and mass distribution in the sample or in a respective partition, optionally (for example, with a deviation greater than a maximum threshold) an iteration back to step S2 especially to measure based on a new collimated approach.

In a fourth step S4 evaluation of measured gamma radiation takes place, in particular for calculating the energy-dependent photopeak efficiency, wherein the fourth step may comprise at least one of the following substeps: evaluation of interactions within the sample ( S4.1 ) for the calculation of energy-dependent photopeak efficiencies, evaluation of interactions in a respective detector ( S4.2 ), Determining the solid angle between sample and detector ( S4.3 ), Determining photopeak efficiencies, in particular (initial) photopeak efficiencies ( S4.4 ). In a fourth regulation point R4 In particular, a transfer and review of interim results can take place.

In a fifth step S5 the mass of at least one element is determined, wherein the fifth step may comprise at least one of the following sub-steps: determining at least one element mass or determining element mass ratios ( S5.1 ), Determining at least one cross-section ( S5.2 ), in particular either from step S1 or step S4 , In a fifth regulation point R5 In particular, a transfer and review of interim results can take place. This can be the control point R5 comprise a plausibility check, in particular a comparison or comparison of quantified elemental masses and the total mass of the sample.

In a sixth step S6 the calculation of the neutronics is carried out, wherein the sixth step may comprise at least one of the following sub-steps: evaluation of interactions, in particular neutron interactions within the sample ( S6.1 ), in particular by diffusion approximation, evaluation of a neutron spectrum ( S6.2 ), Evaluation of a neutron flux ( S6.3 ), in particular by diffusion approximation.

control point R1 to R5 may each include an optional feedback (control loop) to the previous step, in particular as part of a review of a user input or an intermediate result. The steps S4 to S6 can be carried out iteratively, independently of the individual control points, in particular continuously during the evaluation of gamma radiation from continuously irradiated samples. The iteration is aborted if the element mass to be determined no longer changes, or at least no longer changes significantly, for example, from a predefinable threshold value for a difference.

4 shows the field of view of a respective detector 16A . 16B the example of a sliced and circle segments P1 . P2 . pn partitioned cylindrical sample 1 , The field of view of the respective detector 16A . 16B it does not necessarily have to match or be aligned with a respective partition. In the evaluation, it can be taken into account to what proportions an adjacent partition lies within the field of view of the respective detector and is to be evaluated simultaneously with it or eliminated. There are 12 partitions in each height position. The entire sample can then be analyzed by six rotations and by the corresponding number of translational height displacement steps (in this case five levels, ie four displacement steps in the z direction). Each partition is, for example, irradiated and measured over a period of a few seconds to minutes.

In 5 is a device 10 shown at which the sample carrier 14 in height by a considerable distance (arrow with dotted line) can be moved upwards. The sample chamber 15 is through material M3 delimited what material M3 together with the sample 1 in an air-filled cavity above the sample 1 can be relocated. The turning and lifting device 18 is by means of a coupling comprising a shaft 18.1 with the sample carrier 14 but apart from being located outside the neutron shield and sealed off from the sample chamber. This can ensure that the neutrons do not reach the turning and lifting device. A Wegsamkeit the neutrons to the rotating and lifting device is prevented. The material in which the shaft 18.1 is guided, is preferably graphite. As in 5 indicated, in a graphite block a Durchfürhung for the shaft 18.1 be provided. The rotary and lifting device 18 is preferably exclusively by means of the shaft 18.1 with the sample carrier 14 connected. The neutron shield is then broken only by the wave. The rotating and lifting device is arranged in a sealed manner behind the effective neutron shielding.

LIST OF REFERENCE NUMBERS

1

sample

10

Device for multi-element analysis based on neutron activation

11

neutron generator

11.1

Neutron source or neutron source point

12

moderation chamber

14

sample carrier

15

sample chamber

16

detector unit

16A, 16B

single detector

16.1

Crystal of a single detector

16.2

Detektorendkappe

16.3

crystal holder

17

collimator

18

Turning and lifting device

18.1

Coupling, in particular shaft

19

shielding

20

control device

21

data storage

22

nuclear physical database

23

Scanner / mask

24

Transmission measuring unit

25

camera unit

27

weighing unit

30

A computer program product

A16

Visual axis of the detector

M1

Material 1, in particular borated polyethylene and / or concrete

M2

Material 2, in particular lead and / or bismuth

M3

Material 3, in particular graphite

M4

Material or medium 4, in particular air

M6

Material 6, in particular lithium polyethylene and / or lithium silicone

M7

Material 7, in particular aluminum carbon fiber reinforced plastic

M8

Material 8, in particular copper or plastic

M10

Material 10, in particular germanium

P1, P2, Pn

Partitions of the sample

R1

first regulatory point

R2

second regulatory point

R3

third regulation point

R4

fourth control point

R5

fifth control point

S1

first step, in particular generating neutrons and irradiation with neutrons

S1.1

Adjust (control or regulate) the neutron source strength

S1.2

presentation

S1.3

simulative calculation of the neutron spectrum

S1.4

simulative calculation of the neutron flux

S2

second step, in particular sample specification and measurement

S2.1

Detecting sample mass and / or sample geometry and / or transmission measurement

S2.2

collimation

S2.3

Detecting or adjusting sample partitioning

S2.4

Relocation / positioning of the sample, in particular by translation and / or rotation

S3

third step, in particular detection / measurement of emitted gamma radiation and evaluation of measured gamma radiation

S3.1

Detection / measurement of gamma radiation or evaluation of the gamma spectrum

S3.2

Elements / Peak Identification

S3.3

interference analysis

S3.4

Evaluation of peaks, in particular with regard to peak area and background

S4

fourth step, in particular evaluation of measured gamma radiation

S4.1

Evaluate interactions within the sample

S4.2

Evaluation of interactions in the detector

S4.3

Determining the solid angle between sample and detector

S4.4

Determining photopeak efficiencies, in particular initial photopeak efficiencies

S5

fifth step, in particular determining at least one element, in particular the mass

S5.1

Determining at least one element mass or element mass ratios

S5.2

Determining at least one cross-section

S6

sixth step, in particular calculation of neutronics

S6.1

Evaluate interactions within the sample

S6.2

Evaluation of a neutron spectrum

S6.3

Evaluation of a neutron flux

v1

first variable / parameter, in particular manually entered, in particular neutron source strength

v2

second variable / parameter, in particular manually entered, in particular sample geometry

v3

third variable / parameter, in particular manually entered, in particular sample mass

x

longitudinal axis

y

transverse axis

z

vertical axis or vertical axis

QUOTES INCLUDE IN THE DESCRIPTION

This list of the documents listed by the applicant has been generated automatically and is included solely for the better information of the reader. The list is not part of the German patent or utility model application. The DPMA assumes no liability for any errors or omissions.

Cited patent literature

• WO 2012/010162 A1 [0004]
• DE 102010031844 A1 [0004]
• EP 1882929 B1 [0005]
• WO 0107888 A2 [0005]
• EP 0493545 B1 [0005]
• DE 102007029778 B4 [0005]

Claims (16)

Method for multi-element analysis based on neutron activation, comprising the steps of: generating fast neutrons with energy in the range of 10keV to 20MeV; Irradiating a sample (1) with the neutrons; Measuring the gamma radiation emitted from the irradiated sample to determine at least one element of the sample; characterized in that the irradiation of the sample is carried out unpulsed in a continuous manner, wherein the measuring takes place during the irradiation. Method according to Claim 1 wherein the irradiation or the irradiation and measuring takes place over a period of at least one millisecond or at least one second; and / or wherein measurement of radiation emitted in response to the irradiation during irradiation occurs in a time window of less than 5 microseconds; and / or wherein neutrons are generated with the neutron energy value 2.45 MeV or with at least one neutron energy value from the following group: 2.45 MeV, 14.1 MeV; and / or wherein neutrons are generated with a neutron energy having at least one value in the energy range 10keV to 20MeV, in particular 10keV to 10MeV; and / or wherein neutrons are generated with a maximum neutron energy of 10MeV. Method according to one of the preceding claims, wherein for determining at least one element at least prompt or both prompt and delayed gamma radiation from continuous neutron irradiation is measured and evaluated; and / or wherein for determining at least one element at least temporarily only delayed gamma radiation from continuous neutron irradiation is measured and evaluated; and / or wherein the sample (1), in particular individual partitions (P1, P2, Pn) of the sample are measured collimated, in particular by means of at least one detector (16A) or by means of a plurality of detectors (16A, 16B) with specific regarding the geometry of the partition collimated field of view; and / or wherein the measuring, determining and / or evaluating takes place individually with regard to individual partitions (P1, P2, Pn) of the sample, which partitions are predefined or can be predetermined manually or automatically, in particular by collimation. Method according to one of the preceding claims, wherein for determining at least one element, the gamma radiation emitted from the sample (1) is measured in an energy-resolved manner, in particular by determining photopeak count rates, wherein the determining comprises an energy-resolved evaluation of the measured gamma radiation according to at least one gamma spectrum, in particular according to one gamma spectrum detected by a respective detector (16A, 16B); and / or wherein the measuring / evaluating comprises an energy-resolved measurement / evaluation of the intensity of the gamma radiation emitted from the sample (1). The method of claim 1, wherein the determining comprises evaluating the measured gamma radiation, wherein the evaluating comprises: correlating at least one photopeak of a count rate energy representation based on its energy with an element of the sample; and / or wherein the evaluating further comprises: quantifying the mass fraction of at least one element of the sample, in particular by evaluating the fraction of at least one element contained in the sample after deducting a background signal from the net area of a photopeak which the element comprises in a count rate Causes energy representation; and / or wherein the sample (1) is divided into partitions (P1, P2, Pn) and the emitted gamma radiation is measured and evaluated with respect to a respective partition. Method according to one of the preceding claims, wherein the determining comprises an evaluation of the measured gamma radiation, wherein the evaluation is based on the assumption of a homogeneous mass and / or element distribution in the sample (1), in particular a homogeneous mass and / or element distribution in at least one of a plurality of partitions (P1, P2, Pn) of the sample; and / or wherein the determining comprises evaluating the measured gamma radiation, the evaluating comprising: calculating the neutron flux within the sample (1), in particular within a respective partition (P1, P2, Pn) of the sample, based on a diffusion approximation of the linear Boltzmann Equation, in particular spatially and / or energy-resolved, in particular based on the following relationship: $$\mathrm{\Phi }\left(\underset{\_}{x}.e_n\right)={\displaystyle {\int }_{S^2}\mathrm{\Psi }\left(\underset{\_}{x}.e_n.\Omega \right)}d\Omega ;$$

and / or calculating the neutron spectrum within the sample (1), in particular within a respective partition (P1, Pn, Pn) of the sample, in particular spatially and / or energy-resolved, in particular based on the following relationship: $$\mathrm{\Phi }\left(\underset{\_}{x}.e_n\right)={\displaystyle {\int }_{S^2}\mathrm{\Psi }\left(\underset{\_}{x}.e_n.\Omega \right)}d\Omega ,$$

The method of claim 1, wherein the determining comprises evaluating the measured gamma radiation, wherein the evaluating comprises: calculating energy-dependent photopeak efficiencies as well as neutron flux and neutron spectrum within the sample (1) or within a single partition (P1, P2, Pn ) of the sample, in particular calculating neutron flux and neutron spectrum by an approximate method, each based on the following relationship: $$\mathrm{\Phi }\left(\underset{\_}{x}.e_n\right)={\displaystyle {\int }_{S^2}\mathrm{\Psi }\left(\underset{\_}{x}.e_n.\Omega \right)}d\Omega ;$$

and / or wherein the determining comprises evaluating the measured gamma radiation, wherein in evaluating a plurality of gamma energies of at least one element in a respective partition (P1, P2, Pn) of the sample (1) in quantifying the mass fraction of a respective element of the sample respective partition based on the following context: $${\left(P_R\right)}_{e_{\mathrm{\gamma }}}^i=\frac{N_A}{M}\cdot {\displaystyle \underset{k=1}{\overset{N}{\Sigma }}{m}_k}\cdot {\epsilon }_{e_{\mathrm{\gamma }}}^{i\leftarrow k}\cdot {\sigma }_{e_{\mathrm{\gamma }}}^{ik}\cdot {\mathrm{\Phi }}_k^i,$$

Method according to one of the preceding claims, wherein the method is carried out based on the input quantities neutron source strength, sample geometry and sample mass, in particular exclusively based on these three input variables; and / or wherein the method is carried out automatically, in particular by evaluating the measured gamma radiation based on apart from the three parameters neutron source intensity in the irradiation, sample geometry and sample mass, exclusively numerically determined parameters; and / or wherein at least one measurement from the following group is carried out to characterize the sample: transmission measurement, sample weighing, optical detection of the sample geometry; and / or wherein the method is carried out iteratively, in particular in each case with regard to individual elements or with respect to the complete composition of the sample (1) or with respect to individual partitions (P1, P2, Pn) of the sample. Use of a detector unit (16) with at least one detector (16A, 16B) in the multi-element analysis of a sample (1) based on neutron activation arranged to continuously measure both prompt and delayed gamma radiation due to continuous irradiation of the sample with neutrons, the gamma radiation at least partly also continuously and simultaneously for continuous irradiation, in particular use of the detector unit (16) with a plurality of detectors (16A, 16B) respectively collimated or partition collimated or adjustably collimable with respect to at least one partition (P1, P2, Pn) or at least one predefinable geometry of a partition, preferably with a collimator (17) of lead or bismuth. Control means (20) arranged to drive at least one neutron generator (11) of a device for multi-element analysis based on neutron activation, wherein the neutron generator (11) is adapted to generate fast neutrons with energy in the range of 10keV to 20MeV, the controller (20) being arranged is for driving the neutron generator to generate neutrons and irradiating a sample (1) in an unpulsed continuous manner, and wherein the controller is further adapted to drive at least one detector (16, 16A, 16B) for continuously and / or temporarily measuring out the sample or a single partition (P1, P2, Pn) of the sample emitted gamma radiation simultaneously to the irradiation, in particular adapted for controlling a method according to one of the preceding method claims. A computer program product (30) for multi-element analysis based on neutron activation, and arranged to determine at least one element of an unpulsed neutron continuously irradiated sample (1) by evaluating gamma radiation emitted from the sample, namely prompt and / or delayed gamma radiation based on energy-dependent photopeak efficiencies and neutron flux and neutron spectrum within the sample or within a single partition (P1, P2, Pn) of the sample, in particular adapted to evaluate partition collimated measured gamma radiation based on the following relationship: $${\left(P_R\right)}_{e_{\mathrm{\gamma }}}^i=\frac{N_A}{M}\cdot {\displaystyle \underset{k=1}{\overset{N}{\Sigma }}{m}_k}\cdot {\epsilon }_{e_{\mathrm{\gamma }}}^{i\leftarrow k}\cdot {\sigma }_{e_{\mathrm{\gamma }}}^{ik}\cdot {\mathrm{\Phi }}_k^i,$$

A device (10) for multi-element analysis based on neutron activation, comprising: a neutron generator (11) adapted to generate fast neutrons; a sample chamber (15) and a sample carrier (14) disposed therein; a detector unit (16) having at least one detector (16A, 16B) arranged to measure gamma radiation emitted from an irradiated sample to determine at least one element of the sample; characterized in that the device (10) is arranged for unpulsed continuous irradiation of a sample (1) and is adapted to measure prompt and / or delayed gamma radiation emitted by the irradiated sample during irradiation, in particular adapted to perform a method according to one of the preceding method claims. Device according to the preceding device claim, wherein the device comprises at least one component weakening a background signal of the device from the following group: at least one collimator (17) which limits the field of view of the detector to the sample, in particular to a partition of the sample, preferably a lead collimator or bismuth, and / or a graphite moderator chamber (12), and / or a borated polyethylene shield (19), and / or a sample chamber (15) and / or a sample carrier (14) each at least partially graphite or fully fluorinated Plastics or beryllium. Device according to one of the preceding device claims, wherein the neutron generator (11) comprises a neutron source (11.1) arranged for the fusion of deuterons, in particular with deuterium gas as fuel. Apparatus according to any one of the preceding apparatus claims, further comprising a computer program product (30) or data memory (21) therewith, wherein the computer program product is arranged to determine at least one element of the sample (1) by evaluating the measured gamma radiation based on energy dependent photopeak efficiencies and neutron flux and neutron spectrum within the sample or within a single partition (P1, P2, Pn) of the sample; and / or further with a rotating and / or lifting device (18) arranged for translational and / or rotational displacement of a / of the sample carrier (14) or the sample, preferably one from a / the sample chamber (15) of the device decoupled rotary and / or lifting device; and / or furthermore comprising at least two detectors (16A, 16B), in particular in a symmetrical arrangement relative to the neutron generator (11) and / or relative to at least one neutron source (11.1) of the device; and / or further with a control device (20) arranged for automated continuous irradiation and / or set up for controlling an automated measurement of continuously applied neutron irradiation during irradiation, in particular adapted for iterative automated evaluation of the emitted and measured gamma radiation based on the three Parameters Neutron source intensity during irradiation, sample geometry and sample mass, exclusively numerically determined or read by the controller parameters. Use of a neutron generator (11) adapted for the fusion of deuterons, in particular with deuterium gas as fuel, for generating fast neutrons for multi-element analysis of a sample based on neutron activation, in particular for continuous, unpulsed irradiation of the sample, in particular in a method according to one of the preceding method claims.

Images